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Sample records for bwr nuclear plant

  1. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  2. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  3. BWR 90+ - Nuclear power plant for 21st century

    International Nuclear Information System (INIS)

    BWR 90+ is a boiling water reactor, based on the previous models BWR90 and BWR75, and on the operational experiences gained with six reactors of the previous generation. The development work started in 1994 in co-operation with Teollisuuden Voima Oy (TVO). At present all the boiling water reactor owners participate the cooperation. The objectives of the development were: (1) to develop a boiling water reactor of competitive price level and short construction time, and which meets the latest safety requirements, (2) to itemize the technologies improving the security and competitivity of present plants, and (3) to maintain the expertise of the personnel of the companies participating the development work, and improving the BWR- technology. High power output and short construction time reduce the power generation costs. Large amount of fuel assemblies leads to higher safety margins. Reduction of scram groups from 18 to 16 reduces the amount of components, the assembly space and costs. The reactor technical data is as follows: Thermal power output 4250 MWth; electric power output 1500 MWe, construction time 1500 days, costs 1500 pounds/kWe, no. of fuel assemblies 872, no. of scram groups 16, turbines 1, the capacity factor 90% and the duration of service outage 3 weeks. Specific features of BWR90+ are: short construction time and low costs, risk for connection between wet and dry spaces has been minimized, reactor core remains covered by water during loss-of-coolant accident caused by fuel replacement, Passive collection and cooling of core-melt inside the containment, the containment is not the first wall against the spreading of core-melt, steam explosions and core- concrete interactions have low probabilities, high gas- volume of wet-space reduces the pressure increase during a severe accident, filter-equipped gas removal system forms the final overpressure shield, the containment is cylindrical, and the plant is equipped with digital instrumentation and control

  4. Water chemistry management of nuclear power plant. Water chemistry management of BWR plant

    International Nuclear Information System (INIS)

    There are two kinds of nuclear power plants such as Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) in Japan. In this paper, a water chemistry management of BWR plant is explained. BWR plant makes steam produced in the reactor send to the turbine and produce power, then condensate in the main condenser and use again as feed water. The objects of water chemistry management of BWR are security of good conditions of fuel and structure materials and reduction of the dose equivalent and the radioactive waste. The volume of coolant depends on the temperature change, the concentration of boric acid for neutron absorber, lithium hydroxide for pH control and hydrogen gas for corrosion are controlled. Impurity metals in water of reactor are removed by the condensate demineralizer. The concentration of boron and lithium is controlled from 0 to 4000 ppm and from 0.2 to 2.2 ppm, respectively. On water chemistry technologies for dose reduction, oxygen injection into feed water and control operation of rate of Ni/Fe are explained. On the technologies for preventive maintenance, degassing operation of reactor and hydrogen injection into feed water are described. (S.Y.)

  5. The stability analysis of LAPUR6 for Chinshan BWR nuclear power plant

    International Nuclear Information System (INIS)

    Chinshan Nuclear Power Plant (NPP) is the first NPP in Taiwan which is a BWR/4 plant. The original rated power for each unit was 1775 MWt. After the project of measurement uncertainty recovery (MUR) for Chinshan NPP, the operating power is 1805 MWt now. The Chinshan NPP Unit 2 cycle 23 stability analyses were performed by the LAPUR6 stability analysis methodology. Comparing the LAPUR6 stability analysis results and vendor's results, they are similar. (author)

  6. Development of a dynamic model of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    A dynamic model of a nuclear power plant, including a boiling water reactor, high- and low-pressure turbines, moisture separator, reheater, condenser, feedwater heaters and feedwater pump, was developed. The model is one-dimensional except for the nuclear part of the reactor, which is based on the point kinetics equation, and the condenser model and feedwater pump model. It has been used to study different transients occuring during normal operating conditions and for evaluating the control systems of a BWR nuclear power plant. Particular emphasis was laid on the reactor pressure control system and the recirculation flow control system. (author)

  7. Aging and service wear of control rod drive mechanisms for BWR nuclear plants

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''managing'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to normal wear and aging are the Graphiter seals. The predominant causes of aging for these seals are mechanical wear and thermally induced embrittlement More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are valve seals, discs, seats, stems, packing, and diaphragms. Since CRDM changeout and rebuilding is one of the highest dose, most physically challenging, and complicated maintenance activities routinely accomplished by BWR utilities, this report also highlights recent innovations in CRDM handling equipment and rebuilding tools that have resulted in significant dose reductions to the maintenance crews using them

  8. Dry well cooling systems in BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Purpose: To prevent the damages of pipeways due to salt damages at the surface of control rod drives in BWR type reactors. Constitution: In control rod drives and the lowermost area in the dry well in which surface corrosion and pitching have been resulted by the salt contents in air due to the increase in the humidity accompanying the lowering of the temperature, a blower is disposed to the upstream of the cooling coils and a portion of high temperature air returned to the lower cooler is replaced with a low temperature feed air to increase the feed temperature in the area. Further, by upwardly turning the downwarded feed air drawing port in which cold feed air has so far been descended as it is, the descendance of the cold air is suppressed. As a result, temperature lowering in the driving mechanisms and the lower area can be prevented to obtain a predetermined temperature, whereby the dewing on the surface can be prevented and thereby preventing the occurrence of corrosion and pitching. (Horiuchi, T.)

  9. CAE advanced reactor demonstrators for CANDU, PWR and BWR nuclear power plants

    International Nuclear Information System (INIS)

    CAE, a private Canadian company specializing in full scope flight, industrial, and nuclear plant simulators, will provide a license to IAEA for a suite of nuclear power plant demonstrators. This suite will consist of CANDU, PWR and BWR demonstrators, and will operate on a 486 or higher level PC. The suite of demonstrators will be provided to IAEA at no cost to IAEA. The IAEA has agreed to make the CAE suite of nuclear power plant demonstrators available to all member states at no charge under a sub-license agreement, and to sponsor training courses that will provide basic training on the reactor types covered, and on the operation of the demonstrator suite, to all those who obtain the demonstrator suite. The suite of demonstrators will be available to the IAEA by March 1997. (author)

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

    International Nuclear Information System (INIS)

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

  11. Laguna Verde nuclear power plant: an experience to consider in advanced BWR design

    International Nuclear Information System (INIS)

    Laguna Verde is a BWR 5 containment Mark II. Designed by GE, two external re-circulation loops, each of them having two speed re-circulation pump and a flow control valve to define the drive flow and consequently the total core flow an power control by total core flow. Laguna Verde Design and operational experience has shown some insights to be considering in design for advanced BRW reactors in order to improve the potential of nuclear power plants. NSSS and Balance of plant design, codes used to perform nuclear core design, margins derived from engineering judgment, at the time Laguna Verde designed and constructed had conducted to have a plant with an operational license, generating with a very good performance and availability. Nevertheless, some design characteristics and operational experience have shown that potential improvements or areas of opportunity shall be focused in the advanced BWR design. Computer codes used to design the nuclear core have been evolved relatively fast. The computers are faster and powerful than those used during the design process, also instrumentation and control are becoming part of this amazing technical evolution in the industry. The Laguna Verde experience is the subject to share in this paper. (author)

  12. Operational experience of human-friendly control and instrumentation systems for BWR nuclear power plants

    International Nuclear Information System (INIS)

    In recent BWR nuclear power plants in Japan, an advanced centralized monitoring and control system PODIA (Plant Operation by Displayed Information and Automation), which incorporates many operator aid functions, has been in operation since 1985. Main functions of the PODIA system as a computerized operator aid system are as follows. CRT displays for plant monitoring. Automatic controls and operation guides for plant operation. Stand-by status monitoring for engineered safety features during normal operation. Surveillance test procedure guides for engineered safety features. Integrated alarm display. The effectiveness of these functions have been proved through test and commercial operation. It has been obtained that operators have preferred PODIA much more than conventional monitoring and control systems

  13. BWR type nuclear plant and start-up operation method

    International Nuclear Information System (INIS)

    During start-up operation of a reactor, the concentration of dissolved oxygen, hydrogen peroxide or ions in primary coolants is controlled at a low level. In addition, hydrogen is injected to the primary coolants to form a stable oxide membrane on the surface of nuclear elements in contact with water. Further, the flow rate of the primary coolants to be supplied to a reactor clean-up system is determined to not less than 2.1% of the feedwater flow rate during rated power operation, and the flow rate of the reactor clean-up system is controlled so that the electroconductivity of the primary coolants is not more than 0.15μS/cm at the upstream of the clean-up device of the reactor clean-up system. Then, the oxidation rate is suppressed and the concentration of ions in water intaken to the oxide membrane is controlled by decreasing the oxide membrane forming speed on the surface of a zirconium alloy and controlling the concentration of ions in the primary coolants. The ingredients which form a tetragonal or cubic crystal structure if they are involved into the oxide membrane can be incorporated in the primary coolants during start-up operation by the control of the ion concentration. (N.H.)

  14. Parameter identification of a BWR nuclear power plant model for use in optimal control

    International Nuclear Information System (INIS)

    The problem being considered is the modeling of a nuclear power plant for the development of an optimal control system of the plant. Current system identification concepts, combining input/output information with a-priori structural information are employed. Two of the known parameter identification methods i.e., a least squares method and a maximum likelihood technique, are studied as ways of parameter identification from measurement data. A low order state variable stochastic model of a BWR nuclear power plant is presented as an application of this approach. The model consists of a deterministic and a noise part. The deterministic part is formed by simplified modeling of the major plant dynamic phenomena. The moise part models the effects of input random disturbances to the deterministic part and additive measurement noise. Most of the model parameters are assumed to be initially unknown. They are identified using measurement data records. A detailed high order digital computer simulation is used to simulate plant dynamic behaviour since it is not conceivable for experimentation of this kind to be performed on the real nuclear power plant. The identification task consists in adapting the performance of the simple model to the data acquired from this plant simulation ensuring the applicability of the techniques to measurement data acquired directly from the plant. (orig.)

  15. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde

    International Nuclear Information System (INIS)

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  16. Experimental study for quantative aging evaluation of epoxy liner in BWR nuclear power plant

    International Nuclear Information System (INIS)

    The purpose of this study is an experimental approach to quantitatively evaluate the aging status of epoxy coating onto containment structure in BWR nuclear power plant. Based on accelerated aging experiment for 64 days, adhesion test was performed to evaluate an physical bonding. To compare with adhesion data, both impedance data by UT and data by thermal gravimetric analysis were obtained during experiment. At almost 50% of adhesion force decrease, it was identified that aging phenamena of epoxy such as pine hole, blistering was discovered. Coating to establish aging status of epoxy, relations among three kinds of different data were analyze. By compatibility of these data, physical aging situation of as-built epoxy coating was figured out. The possibility to develop new methodology of time-dependent aging status on epoxy coating was identified

  17. Cooperative control scheme for an HVDC system connected to an isolated BWR nuclear power plant

    International Nuclear Information System (INIS)

    This paper describes a cooperative control system to achieve stable operation of an isolated BWR nuclear plant linked to an HVDC system. In the proposed control system, under normal conditions the power plant is controlled according to the generating power reference and the generator frequency deviation is adjusted by converter power control. Such frequency control is also effective in the case of AC-DC system faults. In addition to the frequency control, an overload control is provided with the HVDC system, where the DC transmission power in the sound poles is increased due to a fault detection signal from the faulty pole. Effects of the above mentioned control systems were studied using digital dynamic programs. The sets of simulation results confirmed that in the case of a DC single pole fault, the plant is able to continue operation without any use of the turbine speed control units even for a restarting failure in the faulty pole. In case of a DC two pole fault, the plant is able to continue operation, being assisted by turbine speed control units when restarting in the faulty poles succeeds. In case of an AC three-line to ground fault near the AC terminal of the converter at the sending or receiving end, the system is able to continue stable operation, being supplemented by the turbine control unit when the faulty section of the AC system is isolated by a main or back-up relaying system

  18. Experiences with monitoring and control of microbiological growth in the standby service water system of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    In 1989, the Unites States Nuclear Regulatory Commission formally recognized the potential for nuclear accidents resulting from microbiological causes. Such causes range from loss of heat transfer due to microbiological fouling, to loss of system integrity caused by microbiologically influenced corrosion (MIC). As a result of these potential problems, monitoring, mitigation, and control procedures must be developed by all regulated plants. In developing a control and mitigation strategy for the standby service water system of a boiling water reactor (BWR) nuclear power plant, numerous monitoring techniques were employed to evaluate effectiveness. This paper describes the monitoring techniques that were evaluated, and those that ultimately proved to be effective

  19. TRACE/FRAPTRAN analysis of Kuosheng (BWR/6) Nuclear Power Plant for the similar Fukushima accident

    International Nuclear Information System (INIS)

    Kuosheng nuclear power plant (NPP) is a BWR/6 type NPP and located on the northern coast of Taiwan. In order to assess Kuosheng NPP’s safety under the similar Fukushima accident conditions, Kuosheng NPP safety analysis TRACE/SNAP/FRAPTRAN models were established in this research. There were three main steps in this study. The first step was the establishment of Kuosheng NPP TRACE/SNAP model. The next step was the transient analysis of Kuosheng NPP TRACE/SNAP model under the similar Fukushima accident or more severe conditions. Besides, in order to confirm the mechanical property and integrity of fuel rods, the third step was FRAPTRAN analysis. Finally, the animation model of Kuosheng NPP was presented by using animation function of SNAP. TRACE and FRAPTRAN results depicted that zirconium-water reaction generated in SBO or SBO + LOCA transient (no water injected case). It may cause a safety issue in fuel rods. However, if the fire water (39 kg/sec) injected to the vessel at 800 sec in SBO + LOCA transient, TRACE and FRAPTRAN results implied that the integrity of fuel rod was kept. (author)

  20. An application of risk-informed evaluation on MOVs and AOVs for Taiwan BWR-type nuclear power plants

    International Nuclear Information System (INIS)

    Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs' and AOVs' safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee

  1. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  2. Decontamination techniques for BWR power generation plant

    International Nuclear Information System (INIS)

    The present report describes various techniques used for decontamination in BWR power generation plants. Objectives and requirements for decontamination in BWR power plants are first discussed focusing on reduction in dose, prevention of spread of contamination, cleaning of work environments, exposure of equipment parts for inspection, re-use of decontaminated resources, and standards for decontamination. Then, the report outlines major physical, chemical and electrochemical decontamination techniques generally used in BWR power generation plants. The physical techniques include suction of deposits in tanks, jet cleaning, particle blast cleaning, ultrasonic cleaning, coating with special paints, and flushing cleaning. The chemical decontamination techniques include the use of organic acids etc. for dissolution of oxidized surface layers and treatment of secondary wastes such as liquids released from primary decontamination processes. Other techniques are used for removal of penetrated contaminants, and soft and hard cladding in and on equipment and piping that are in direct contact with radioactive materials used in nuclear power generation plants. (N.K.)

  3. BWR radiation control: plant demonstration

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  4. Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

    International Nuclear Information System (INIS)

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be 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. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

  5. MELCOR/SNAP analysis of Chinshan (BWR/4) Nuclear Power Plant spent fuel pool for the similar Fukushima accident

    International Nuclear Information System (INIS)

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP event occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool, by using MELCOR 2.1 and SNAP 2.2.7 codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP spent fuel pool (SFP). There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP MELCOR/SNAP model. And the transient analysis under the SFP cooling system failure condition was performed. Besides, in order to study the detailed thermal-hydraulic performance of this transient, TRACE was used in this analysis. CFD data from INER report was used to compare with the results of MELCOR and TRACE. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Besides, the animation model of Chinshan NPP SFP was presented using the animation function of SNAP with MELCOR analysis results. (author)

  6. Near-term improvements for nuclear power plant control room annunciator systems. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rankin, W.L.; Duvernoy, E.G.; Ames, K.R.; Morgenstern, M.H.; Eckenrode, R.J.

    1983-04-01

    This report sets forth a basic design philosophy with its associated functional criteria and design principles for present-day, hard-wired annunciator systems in the control rooms of nuclear power plants. It also presents a variety of annunciator design features that are either necessary for or useful to the implementation of the design philosophy. The information contained in this report is synthesized from an extensive literature review, from inspection and analysis of control room annunciator systems in the nuclear industry and in related industries, and from discussions with a variety of individuals who are knowledgeable about annunciator systems, nuclear plant control rooms, or both. This information should help licensees and license applicants in improving their hard-wired, control room annunciator systems as outlined by NUREG-0700.

  7. Manage I and C obsolescence for GE BWR nuclear power plant life extension

    International Nuclear Information System (INIS)

    Most nuclear power plants in operation were built in the 1970's and 1980's. Parts obsolescence for instrumentation and controls (I and C) has been a continued issue for these plants. Although modern I and C could benefit a nuclear plant in terms of improved reliability, reduced maintenance, and reduced plant vulnerability, I and C upgrades are not common for nuclear plants. This is because the current economic model is such that the assessed benefits for I and C upgrades, especially for the safety related applications, seldom exceeds the modifications cost. Generally, there is no capacity loss associated with I and C problems. The plant staff is usually talented and resourceful enough to maintain the existing system. Any upgrade involving safety related instruments require extensive reviews by the regulatory body, e.g., the Nuclear Regulatory Commission. Cost/benefit analysis that focuses on the immediate pay back and discounts potential long term impact could not justify the cost of I and C upgrade. Consequently, I and C upgrade has been conducted more from a as-needed basis as opposed to being based on a strategic plan. The planning of the I and C upgrades to support plant life extension should be executed in a systematic approach. Six Sigma tools, such as Quality Function Deployment (QFD), can be used to help prioritize I and C upgrades for coming years. An example of employing such a tool would be to first list the criteria or goals for upgrades. This is also termed 'Critical to Quality (CTQ)'. The CTQ should be developed based on inputs from plant staff including engineering, maintenance, and operations. Where applicable, quality assurance, outage management, and plant strategic planning staff should be included. Examples of CTQs are: (a) reliability, (b) ease of maintenance, (c) ease of operation, (d) outage impact, (e) long tem viability, (f) ALARA, and (g) ability to support plant operation (e.g. EPU). The CTQs are then weighted based on the consensus of

  8. Development of a dynamic model of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    A description is given of a one-dimensional steady-state model of a high-pressure steam turbine, a low-pressure steam turbine, a moisture separator, a reheater, a condenser, feedwater heaters and feedwater pump for a nuclear power plant. The model is contained in the program ''TURBPLANT''. The dynamic part of this model is presented in part II of this report. (author)

  9. Development of electrical cable penetration for secondary containment vessel of BWR type nuclear power plants

    International Nuclear Information System (INIS)

    The penetration holes in the walls and floors of the secondary containment vessel of the nuclear power plants must be air-tight, shielded against the radiation, and fire-resistant. At present, the penetration holes are air-tightened with iron plates and sealing material after the cables are laid. However, installation of a number of cables and its sealing work now pose a serious problem in nuclear power plant construction in relation to the installation of reactor system components. The authors have recently developed a method for electric wall penetration in an attempt to solve this problem. This method is provided with prefabricated cable portions for wall penetration, reducing field work, saving labor in wiring work through use of multicore cables, and increasing the reliability of the sealing and caulking work. This wall penetration consists of an iron sleeves to be embedded into the wall, a header-plate, and an assembly of modules in which a specified number of insulated conductors are set up, and furthermore termination boxes are installed on both ends of the penetration holes. This paper deals with the design standard and construction of the wall penetration and the results of tests which were performed under various environmental conditions, which has shown excellent properties, such as sealing quality and electric characteristics, of the wall penetration. (author)

  10. GARLIC-B. A digital code for real-time calculation of the transient behaviour of nodal and global core and plant parameters of BWR nuclear power plants

    International Nuclear Information System (INIS)

    A program description of the code GARLIC-B is given. The code is based on a nonlinear transient model for BWR nuclear power plants which consist of a 3D-core, a top plenum, steam removal and feed water systems and a downcomer with main coolant recirculation pumps. The core is subdivided into a number of superboxes and flow channels with different coolant mass flow rates. Subcooled boiling within these channels has an important reactivity feed back effect and has to be taken also into account. The code computes the local and global core and plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. (orig./GL)

  11. Experience of MOX-fuel operation in the Gundremmingen BWR plant: Nuclear characteristics and in-core fuel management

    International Nuclear Information System (INIS)

    After 4 years of good experience with MOX-fuel operation in the BWR plants Gundremmingen units B and C the number of inserted MOX-FAs will be increased in the future continuously. Until now all MOX-FAs are in good condition. Furthermore calculations and measurements concerning zero power tests and tip measurements are in good agreement as expected: all results lead to the conclusion that MOX-FAs can be calculated with the same precision as uranium-FAs. (author)

  12. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde; Estudio de ruido ambiental en una planta BWR como la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  13. BWR manual control and status monitoring system for nuclear power plant

    International Nuclear Information System (INIS)

    As the unit capacity of nuclear power plants increases, the reduction of cost for construction and the improvement of operation availability have been strongly desired. The control and monitoring system with all solid-state equipments, in which the number of cables is greatly reduced by employing series transmission system as signal transmission, has been developed. The new system consists of the control rod drive control system and the control rod monitoring system. The former is composed of the control rod selector module, the control rod drive control panel, distributor module and the transponder. The latter comprises the position sensing probe, the control rod position information panel, whole core display, four control rods display and display memory module. The system operates in 3 modes of time-sharing control, scan, and automatic test. Compared with conventional systems, the new system is improved in three points. (1) The number of cables has been reduced to about 45% of that of conventional system, (2) system availability has been improved by adopting duplex circuits and on-line automatic test function, and (3) dimensions of the control panel has been reduced to about 35% of that of conventional system. In the control rod drive control system, failure diagnosing system is provided, which indicates the diagnosed results or failure positions on a plasma display by utilizing a micro-computer, and almost eliminates the necessity of searching failure positions in the complicated interior of equipments. (Wakatsuki, Y.)

  14. Plant operation performance improvements of the General Electric (GE) boiling water reactors (BWR'S)

    International Nuclear Information System (INIS)

    This paper summarizes some of the plant operation performance improvement techniques developed by the General Electric Company Nuclear Energy Business Operation for the General Electric Boiling Water Reactors (GE BWR's). Through the use of both thermal and plant hardware operating margins, substantial additional flexibility in plant operation can be achieved resulting in significant improvements in plant capacity and availability factor and potential fuel cycle economics for the currently operating or requisition GE BWR plants. This list of techniques includes expanding the BWR thermal power/moderator flow operating domain to the maximum achievable region, operation with a single recirculation loop out of service and operation at rated thermal power with reduced feedwater temperatures. These plant improvements and operating techniques can potentially increase plant capacity factor by 1% to 2% and provide additional fuel cycle economics savings to the GE BWR's owners

  15. Development of a rapid prediction technique for the radioactive release rate in nuclear emergency preparedness. Comparison calculation of radioactivity release for BWR plants

    International Nuclear Information System (INIS)

    When a nuclear disaster occurs in Japan, execution of public protection measures are required before or immediately after (i.e. in the early stage of) radioactive material releases. In particular, in order to carry out sequential protection measures in the urgent protective action planning zone (UPZ), it is important to predict the released amounts of radioactive materials and their influence on the environment. However, in the Fukushima Dai-ich Nuclear Power Plant accident, the emergency response support system (ERSS) and the system for prediction of environmental emergency dose information (SPEEDI) which are used for public protection did not enough achieve their function. Therefore, a quick prediction technique (the technique completed the prediction calculation within 3 minutes, including data input time) was developed using limited information about the amounts of radioactive materials that might be released during accident progression in order to predict appropriate public protection measures at an early stage and to avoid excessive dependence on a complex calculation code. It has been applied to typical PWR plants. Now, in this paper, we extended application of this technique to BWR plants, and we verified that the prediction results of the released total amount of radioactive materials were almost greater than the analysis result using the Incident Progress Prediction System based on the severe accident analysis code MAAP. (author)

  16. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  17. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  18. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  19. BWR type nuclear power plant and operation method therefor and method of forming oxide membrane on the surface of the constitutional member in contact with water

    International Nuclear Information System (INIS)

    In a BWR type nuclear power plant, an oxide membrane is formed on the surface of the constitutional members of a reactor primary system to be in contact with water while keeping the reactor water at a pH of 7.5 or less based on a room temperature and keeping a temperature of reactor water at 250degC or higher for 250 hours or more and then adding alkaline water to control the pH within a range of from 7.5 to 9.0 based on the room temperature and keeping the reactor water temperature to 250degC or higher for 100 hours or more. This process is conducted during the reactor shut down state and during the operation period from the time of the reactor shut down state to the time of the rated power operation state of the electric power generator. Then, a corrosion resistant oxide membrane with less involvement of radioactive ions can be formed, thereby enabling to improve corrosion resistance of nuclear fuel elements and suppressing the dose rate on the surface of pipelines of a primary coolant system, accordingly, operator's radiation dose rate can be reduced upon periodical inspection. (N.H.)

  20. Qualification of the TH-3DNK coupled code RELAP5/PARCSV2.7 against a BWR unstable point at peach bottom nuclear power plant

    International Nuclear Information System (INIS)

    In this work, three dimensional time domain BWR stability analysis were performed on a new analysis point (PTUPV), which is inside the exclusion region with a core mass flow of 4660.1 kg/s (34% of the core rated mass flow) and total reactor power of 1997.8 MW (60.7 of the core rated reactor power), using the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by the control rod movement as it is usual performed in Nuclear Power Plants. The transient starts with the control rod movement shown. The control rods move in 6 seconds; at the end of the movement the majority of the banks are completely withdrawn and only the bank 7 is almost completely inserted. The purpose of this study is to qualify this coupled code against this kind of 3D complex accidents that take place inside the core. With the aim to make a more careful analysis of the instability recognized in the RELAP5/PARCS simulation, using the nuclear cross-sections provided by the transient calculation performed with the coupled codes, a number of analyses with the nodal modal code VALKIN has also been carried out.

  1. Validation of LANCR01/AETNA01 BWR code package against FUBILA MOX experiments and Fukushima Daiichi Nuclear Power Plant Unit 3 MOX core

    International Nuclear Information System (INIS)

    LANCR01 assembly code and AETNA01 core simulator are the advanced BWR package developed by Global Nuclear Fuel (GNF). In order to establish the applicability of the package to the plutonium containing mixed oxide (MOX) fuel, validation tests have been conducted against BASALA, FUBILA MOX critical experiments and the operational data from Fukushima Daiichi Nuclear Power Plant Unit 3 (1F3) with MOX assemblies loaded at cycle 25. For BASALA and FUBILA, critical eigenvalue and the pin-by-pin fission rate distribution by LANCR01 were compared with the experimental data. For 1F3, AETNA01 predictions with LANCR01 assembly cross sections were compared with the measured control blade worth and the moderator temperature coefficient in the reactor physics tests, as well as the cold/hot eigenvalues and the in-core instrument readings during the operation. It is concluded that LANCR01/AETNA01 system has a comparable accuracy for the MOX cores with that for the uranium cores. (author)

  2. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

    International Nuclear Information System (INIS)

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize 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. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

  3. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

    International Nuclear Information System (INIS)

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating maintenance or manufacturing errors) can jeopardize 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 ware out of components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of heavy water moderated reactors (HWRs), boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

  4. Radiation source term reduction in BWR plants

    International Nuclear Information System (INIS)

    This series of slides presents: the collective radiation exposures at US and European BWRs; the European experience with source term reduction measures (normal water chemistry - NWC): zinc addition, stellite replacement, full system decontamination; the effects of evolving water chemistries/US experience. The conclusions are summarized as follows: worldwide reduction of collective radiation exposures at BWRs by following the ALARA principle; zinc addition proven option for source term reduction for NWC and hydrogen water chemistry (HWC) plants; reducing feedwater iron has been proven to reduce dose rates - as operational observations in the US indicate; optimized feedwater iron is very important for fuel performance under all modes of water chemistry (HWC, Zn, and noble metal chemical addition (NMCA)); minimize 59Co sources/stellite, follow the ALARA principle; full system decontamination (FSD) plus zinc injection is an attractive option for reducing reactor coolant system (RCS) dose rates of mature BWR plants

  5. Inspection findings in austenitic RPV internals of German BWR plants and BWRs built in other countries and resulting measures for Isar 1 nuclear power station

    International Nuclear Information System (INIS)

    As visual examinations carried out in autumn 1994 detected cracks in a German BWR plant due to intergranular stress corrosion cracking (IGSCC) in several core shroud components manufactured from 1.4550 steel, precautionary examinations and assessments were performed for all other plants. In accordance with these analyses, it can be stated for Isar 1 that the heat treatment to which the components in question were subjected in the course of manufacture cannot have caused sensitization of the material, and that crack formation due to the damage mechanism primarily identified in the reactor vessel internals at Wuergassen Nuclear Power Station need not be feared. Although the material and corrosion-chemical assessments performed to date did not give any indications for the other crack formation mechanisms that are theoretically relevant for reactor vessel internals (IGSCC due to weld sensitization, IASCC (irradiation assisted stress corrosion cracking)), visual examinations with a limited scope will be carried out with the independant expert's agreement during the scheduled inservice inspections. The fluid-dynamic and structure-mechanical analyses showed that the individual components are subjected only to low loadings, even in the event of accidents, and that the safety objectives shutdown and residual heat removal can be fulfilled even in the case of large postulated cracks. The fracture-mechanics analyses indicated critical through-wall crack lengths which, however, can be promptly and reliably detected during random inservice inspections even when assuming stress corrosion cracking and irradiation-induced low-toughness material conditions. In addition, both the VGB and the Isar 1 plant are pursuing further prophylactic measures such as alternative water chemistry modes and an appropriate repair and replacement concept. (orig.)

  6. Procedures for using expert judgment to estimate human-error probabilities in nuclear power plant operations. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Seaver, D.A.; Stillwell, W.G.

    1983-03-01

    This report describes and evaluates several procedures for using expert judgment to estimate human-error probabilities (HEPs) in nuclear power plant operations. These HEPs are currently needed for several purposes, particularly for probabilistic risk assessments. Data do not exist for estimating these HEPs, so expert judgment can provide these estimates in a timely manner. Five judgmental procedures are described here: paired comparisons, ranking and rating, direct numerical estimation, indirect numerical estimation and multiattribute utility measurement. These procedures are evaluated in terms of several criteria: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Situational constraints such as the number of experts available, the number of HEPs to be estimated, the time available, the location of the experts, and the resources available are discussed in regard to their implications for selecting a procedure for use.

  7. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    An engineering plant analyzer has been developed at BNL for realistically and accurately simulating transients and severe abnormal events in BWR power plants. Simulations are being carried out routinely with high fidelity, high simulation speed, at low cost and with unsurpassed user convenience. The BNL Plant Analyzer is the only operating facility which (a) simulates more than two orders-of-magnitude faster than the CDC-7600 mainframe computer, (b) is accessible and fully operational in on-line interactive mode, remotely from anywhere in the US, from Europe or the Far East (Korea), via widely available IBM-PC compatible personal computers, standard modems and telephone lines, (c) simulates both slow and rapid transients seven times faster than real-time speed in direct access, and four times faster in remote access modes, (d) achieves high simulation speed without compromising fidelity, and (e) is available to remote access users at the low cost of $160 per hour. The accomplishment of detailed and accurate simulations in complex power plants at high speed and low cost are due chiefly to two reasons. The first reason is the application of five distinct modeling principles [2] which are not employed in any other simulation code. The second, and even more important reason is the utilization of a special-purpose peripheral computer with its 13 task-specific parallel processors

  8. Control method for BWR type power plant

    International Nuclear Information System (INIS)

    The present invention provides a method of controlling a BWR type plant having internal pumps capable of sufficiently utilizing the performance of a whole volume turbine bypass plant to enable stable supply of electric power upon load interruption of power generator thereof. Namely, upon occurrence of load interruption of a power generator or turbine trip, a plurality of internal pumps are tripped simultaneously to abruptly reduce a reactor core flow rate by a predetermined value or more. In this case, a reactor core flow rate abruptly reduction scram signal is prevented. Alternatively, a plurality of internal pumps are tripped simultaneously to abruptly reduce the reactor core flow rate. In this case, a reactor core flow rate abrupt reduction scram set value is changed in order to inhibit the reactor core flow rate abrupt reduction scram signal. With such procedures, upon load interruption of power generator or upon trip of turbine, reactor core flow rate is abruptly reduced by trip of internal pumps for avoiding increase of neutron fluxes due to reactor pressure change. However, since reactor scram is avoided, the operation can be continued upon load interruption of power generator. As a result, performance of whole volume turbine bypass plant can be utilized sufficiently even upon occurrence of load interruption of power generator. (I.S.)

  9. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  10. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  11. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  12. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

  13. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  14. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  15. In-reactor ECP measurements in BWR plants

    International Nuclear Information System (INIS)

    It has previously been confirmed that the initiation and propagation of Stress Corrosion Cracking (SCC) of stainless steel and Ni-based alloy exposed in the primary water of Boiling Water Reactor (BWR) depend on water chemistry. It is known that the corrosion environment is evaluated with Electrochemical Corrosion Potential (ECP) and SCC susceptibility is high when the ECP is high. Then it also has been confirmed that hydrogen injection in feedwater and Noble Metal Chemical Addition (NMCA) are effective to reduce ECP of reactor components. ECP measurements in BWR plants, which are BWR-3 and BWR-4, were performed in Normal Water Chemistry (NWC), Hydrogen Water Chemistry (HWC) and post-NMCA environments to evaluate the mitigation effect of SCC by HWC and NMCA. ECP measurements were conducted in the lower plenum region, bottom head region, below the core plate and near the bottom of the active fuel region, by installing modified LPRM with ECP electrodes, and at bottom head drain line flange location (BHDL). It is confirmed that the ECP are reduced to less than -200mV(SHE) by 0.9 ppm H2 concentration at feedwater in BWR-3 and 1.1 ppm H2 concentration at feedwater in BWR-4, and that the ECP can be reduced to less than -200mV(SHE) by 0.3 ppm or less H2 concentration after NMCA is applied. (author)

  16. Reliability improvement method for BWR power plants

    International Nuclear Information System (INIS)

    The construction of the nuclear power generating facilities in Japan was commenced by the import of technological know-how from the United States, but in ten years since then, they reached the stage of improvement and standardization by the effort for the domestic production and the accumulated technological ability. But the unscheduled stop of operation was not able to avoid centering around the initially imported plants, and it cannot be said that the sufficient rate of operation was attained. In Japan, plant manufacturers deliver the whole installations including nuclear reactors in the lump, and carry out the planning, design, manufacture, construction, periodic inspection and maintenance, accordingly the feedback of the operational results can be made quickly, differing from the U.S. system. As the result, No. 1 plant of the Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc., which was constructed by the domestic technology, has attained about 72% of the average rate of operation in six years, and showed the high reliability of the domestically produced plants. The measures for improving system reliability in system planning and the reliability of machinery and equipments, the method of evaluating the reliability of systems, machines and equipments, the quality of nuclear power generating facilities and the quality assurance, and the management of maintenance in Hitachi Ltd. are explained. (Kako, I.)

  17. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  18. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  19. A BWR power plant simulator for Barsebaeck

    International Nuclear Information System (INIS)

    A computer simulator of a Barsebaeck power plant unit has been developed in cooperation between Sydkraft AB, Lund Institute of Technology, and Risoe National Laboratory. The simulator is of the kind often referred to as a compact simulator, because it involves only a computer with display screens and other input/output devices plus the software needed for calculation and presentation of the plant state as a function of time, and no sort of model of the control room as in large reactor simulators for operator training. The purpose of training courses with the compact simulator is to give students a better understanding of the behaviour of the power plant under transient conditions by displaying variables, e.g. pressures, temperatures, reactivity, nuclear power, as functions of time, thereby showing the interactions between different parts of the plant during the transient and the influence of a number of possible operator actions. The present paper describes the Barsebaeck compact simulator with the emphasis on the software developed at Risoe National Laboratory. The Risoe work comprises the programming of the dynamic plant model, in the form of a number of Fortran subroutines containing the physical description of the power plant. (author)

  20. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m3/hour with cost US$ 0.58/m3. The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  1. ATRIUMTM Fuel - Continuous Upgrading for High Duty BWR Plants

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to Boiling Water Reactors worldwide, representing today more than 60 000 fuel assemblies. Since first delivered in 1992, ATRIUMTM10 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. Among them, the latest versions are ATRIUMTM 10XP and ATRIUMTM 10XM fuel assemblies which have been delivered to several utilities worldwide. During six years of operation experience reaching a maximum fuel assembly burnup of 66 MWd/kgU, no fuel failure of ATRIUMTM 10XP/XM occurred. Regular upgrading of the fuel assemblies' reliability and performance has been made possible thanks to AREVA NP's continuous improvement process and the 'Zero tolerance for failure' program. In this frame, the in-core behavior follow-up, manufacturing experience feedback and customer expectations are the bases for setting improvement management objectives. As an example, most fuel rod failures observed in the past years resulted from debris fretting and Pellet Cladding Interaction (PCI) generally caused by Missing Pellet Surface. To address these issues, the development of the Improved FUELGUARDTM debris filter was initiated and completed while implementation of chamfered pellets and Cr doped fuel will address PCI aspects. In the case of fuel channel bow issue, efforts to ensure dimensional stability at high burnup levels and under challenging corrosion environments have been done resulting in material recommendations and process developments. All the described solutions will strongly support the INPO goal of 'Zero fuel failures by 2010'. In a longer perspective, the significant trend in nuclear fuel operation is to increase further the discharge burnup and/or to increase the reactor power output. In the majority of nuclear power plants worldwide, strong efforts in power up-rating were made and are still ongoing. Most

  2. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  3. Impact of Plant Noise on BWR Stability Analyses

    International Nuclear Information System (INIS)

    A small amount of thermal-hydraulic noise is present in a boiling water reactor (BWR). The noise originates primarily from minor fluctuations in fluid flow and pressure distribution in the recirculation system of the BWR and manifests itself as a small fluctuation on the order of 1-2% for the average power range monitors (APRM) during normal operation. A larger noise level is observed for single-loop operation than for two-loop operation. This noise has an impact on the stability performance of the BWR. This is particularly the case when the noise contains a significant component at the resonant frequency for BWR instabilities, which is typically on the order of 0.5 Hz. For a pump trip event that can lead to instability, the noise will impact the growth rate of the reactor instability. The initial magnitude of the oscillations will be larger as the decay ratio increases above unity. For operation at low flow, such as for minimum pump speed or single loop operation where the decay ratio is larger, the impact of noise could lead to small oscillations at the resonant frequency for the APRM signals. The impact of noise on BWR instabilities is analyzed with the TRACG code. TRACG consists of a multi dimensional two-fluid thermal hydraulics model and the three-dimensional kinetics model consistent with the GE 3D core simulator, PANACEA. TRACG models the reactor primary system and has been extensively qualified against test data and BWR plant data. Thermal hydraulic instability test data, as well as data from BWR instability events and tests, have been used extensively in this qualification. This paper demonstrates the impact of noise on BWR stability response for events leading to instability, such as pump trip events, as well as operation at low core flow due to single loop operation. The impact is illustrated through sensitivity studies with the TRACG code and by comparison to plant data. The impact of reactor noise on the performance of the instability detection system

  4. Nuclear transmutation characteristics of reduced moderation BWR (Thesis)

    International Nuclear Information System (INIS)

    In the present thesis, the nuclear transmutation characteristics of reduced moderation BWR, which decides the spent fuel characteristics and its safety in its nuclear fuel cycle, were investigated and compared with other types of reactors. The major conclusions were obtained as follows: The decay heat and radioactivity from FPs increases in fuel burn-up. However, they which normalized with burn-up are small for the reactor with low specific power and long operation period due to the decay during the long operation period. Breeder type of reduced moderation BWR shows low decay heat and radioactivity from FPs because of the long operation period approximately 3000 days which realized by the high conversion ratio. That also shows low decay heat and radioactivity from actinide nuclides due to the hard spectrum. MA recycling reactor of high conversion type of reduced moderation BWR was designed. The neptunium, which has large impact for environmental burden from the viewpoint of nuclide transport analysis, can be incinerated approximately 40% of loaded inventory which corresponds to 22 units of LWR per year. LLFP (99Tc, 129I, 135Cs) transmutation by breeder type of reduced moderation BWR was estimated. As a result, the support factor cannot be lower than unity for each LLFP nuclides. In other words, the reduced moderation BWR cannot reduce LLFP because the LLFP target cannot be loaded inner of the reactor core due to the small margin of core specification. It is expected that these results and the characteristics of other types of reactor shown in the present study benefit the discussion for various nuclear fuel cycle options. (author)

  5. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  6. Coupled BWR calculations with the numerical nuclear reactor software system

    International Nuclear Information System (INIS)

    The Numerical Nuclear Reactor (NNR) is a software suite for integrated high-fidelity reactor core simulations including neutronic and thermal-hydraulic feedback. Using solution modules with formulations to reflect the multi-dimensional nature of the system, NNR offers a comprehensive core modeling capability with pin-by-pin representation of fuel assemblies and coolant channels. Originally developed for pressurized water reactors, the NNR analysis capabilities have recently been extended for boiling water reactor (BWR) applications as part of EPRI Fuel Reliability Program. The neutronics methodology is extended to treat non-periodic structure of BWR fuel assemblies, and a new Eulerian two-phase CFD boiling heat transfer model has been integrated with the software system. This paper summarizes the experience with, and results of, the first-of-a-kind coupled calculations as demonstration of a fully-integrated, high-fidelity simulation capability for assessment of margin to crud-induced failure from fuel-duty perspective. (authors)

  7. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  8. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    International Nuclear Information System (INIS)

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise

  9. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  10. BWR radiation control: plant demonstration. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  11. Experimental evaluation of an operator decision aid system for BWR power plants

    International Nuclear Information System (INIS)

    An experiment was carried out to evaluate the effectiveness of an integrated operator decision aid (IODA) system for boiling water reactor (BWR) power plants. The IODA was developed to enhance the operating safety, reliability and quality of plants. It assists an operator's monitoring and diagnostic capabilities under adverse plant situations using computers and color CRT display devices. Three functions - a standby systems management system (SSMS), a disturbance analysis system (DAS), and a post-trip operational guidance (PTOG) system - have been developed as aids corresponding to the operator's various roles in the safe operation of a nuclear power plant. These functions were designed with full consideration of the operator's problem-solving process, and they have been integrated into a comprehensive IODA system for abnormal BWR power plant conditions. To ensure the applicability of the test results to real-life situations, emphasis was placed on establishing realistic test conditions - including a fully equipped experimental control room, a full-scope plant simulator, participation of experienced operating crews, and well-designed test transient scenarios. Experimental data were collected by means of computers and audio-visual devices, and were processed and summarized to facilitate analysis. Qualitative analysis of the test results was performed with emphasis on the decision-making process of an operator. The analysis points to the utility of the IODA as an operator aid, especially in diagnosing adverse plant situations and in formulating strategies for countermeasures. (orig.)

  12. Dynamic safety systems in BWR plant safety systems

    International Nuclear Information System (INIS)

    Dynamic Safety Systems (DSSs) are reactor safety function systems that are functionally controlled using dynamic rather than static processes. All components including software, whose failure could result in a critical safety system failure, are operationally verified by hard-wired components. Dynamic Safety Systems have been enveloped in the United Kingdom by AEA Technology for use in gas cooled reactors. One such system, known as ISAT trademark, is described in this paper. Through use of scenario testing of a DDS emulator on a Boiling Water Reactor plant training simulator described in this paper. Through use of scenario testing of a DSS emulator on a Boiling Water Reactor plant training simulator, it is shown that a DSS can provide a cost effective safety system in BWR power plants

  13. BWR plant analyzer development at BNL [Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    An engineering plant analyzer has been developed at BNL for realistically and accurately simulating transients and severe abnormal events in BWR power plants. Simulations are being carried out routinely with high fidelity, high simulation speed, at low cost and with unsurpassed user convenience. The BNL Plant Analyzer is the only operating facility which (a) simulates more than two orders-of-magnitude faster than the CDC-7600 mainframe computer, (b) is accessible and fully operational in on-line interactive mode, remotely from anywhere in the US, from Europe or the Far East (Korea), via widely available IBM-PC compatible personal computers, standard modems and telephone lines, (c) simulates both slow and rapid transients seven times faster than real-time in direct access, and four times faster in remote access modes, (d) achieves high simulation speed without compromising fidelity, and (e) is available to remote access users at the low cost of $160 per hour

  14. BWR stability analysis with the BNL Engineering Plant Analyzer

    International Nuclear Information System (INIS)

    March 9, 1989 instability at the LaSalle-2 Power Plant and more than ninety related BWR transients have been simulated on the BNL Engineering Plant Analyzer (EPA). Power peaks were found to be potentially seventeen times greater than the rated power, flow reversal occurs momentarily during large power oscillations, the fuel centerline temperature oscillates between 1,030 and 2,090 K, while the cladding temperature oscillates between 560 and 570 K. The Suppression Pool reaches its specified temperature limit either never or in as little as 4.3 minutes, depending on operator actions and transient scenario. Thermohydraulic oscillations occur at low core coolant flow (both Recirculation Pumps tripped), with sharp axial or redial fission power peaking and with partial loss of feedwater preheating while the feedwater is flow kept high to maintain coolant inventory in the vessel. Effects from BOP system were shown to influence reactor stability strongly through dosed-loop resonance feedback. High feedwater flow and low temperature destabilize the reactor. Low feedwater flow restabilizes the reactor, because of steam condensation and feedwater preheating in the downcomer, which reduces effectively the destabilizing core inlet subcooling. The EPA has been found to be capable of analyzing BWR stability '' shown to be effective for scoping calculations and for supporting accident management

  15. Analysis of the integrity of the pressure vessel of the BWR type nuclear reactor

    International Nuclear Information System (INIS)

    The presssure vessel of a BWR type reactor was monitored for cracking during alternating events of its in-service life. The monitoring was to determine criticality of fractures catastrophic fractures and the velocity of fracture propagation. Detected cracks were evaluated as specified in ASME code section XI, of a minimum wall thickness of 2.5% crack growths were compared a) of 1/10 of the critical maximum size and b) at in-service inspection intervals according to ASME recommendations to be established at the Laguna Verde nuclear plant. Finally conclusions are made and discussed. (author)

  16. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices III and IV. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    The items listed below summarize the detail sections which follow: a listing of definitions and a discussion of the general treatment of data within the random variable approach as utilized by the study; a tabulation of the assessed data base containing failure classifications, final assessed ranges utilized in quantification and reference source values considered in determining the ranges; a discussion of nuclear power plant experience that was used to validate the data assessment by testing its applicability as well as to check on the adequacy of the model to incorporate typical real incidents; an expanded presentation of the data assessment giving information on applicability considerations; a discussion of test and maintenance data including comparisons of models with experience data; and special topics, including assessments required for the initiating event probabilities and human error data and modeling.

  17. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  18. Example of a practical demonstration of in-plant inspection methodology (BWR pipework); the demonstration process

    International Nuclear Information System (INIS)

    Following the discovery of widespread cracking in the recirculation piping system at the Nine Mile Point Unit 1 plant, the US Nuclear Regulatory Commission (NRC) made the requirement that the operating BWR plants demonstrate the effectiveness of their ultrasonic testing procedures in detecting intergranular stress corrosion cracks (IGSCC) in large diameter recirculation system pipes. These requirements were issued in the form of bulletins 'IEB 82-03' in October of 1982 and 'IEB 83-02' in March of 1983. EPRI, through the DPRI NDE Centre, took the lead in establishing a program so that the 24 affected BWR plants (nine addressed by IEB 82-03 and 15 addressed by IEB 83-02) could satisfy the NRC requirements with minimum impact on plant schedules. This meant that not only the preparation activities had to be conducted rapidly, but the demonstration process for each team could not last longer than a day, which in itself restricted the size of the test set. This paper describes the elements of the demonstration process and presents a summary of the performances of the inspection teams who took part in the IEB 82-03 and IEB 83-02 activities. (author)

  19. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR

    International Nuclear Information System (INIS)

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  20. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    International Nuclear Information System (INIS)

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  1. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  2. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  3. Alpha-nuclides in nuclear power plants

    International Nuclear Information System (INIS)

    The behaviour of alpha-nuclides in nuclear power plants is subject of the investigations presented. The source of alpha-nuclides is a contamination with fissile material (so called tramp uranium or tramp fuel) which deposits on fuel rod surfaces and leads to the build-up of transuranium nuclides. The determination of a defect situation with fuel release as well as the quantification of the fissile material contamination background is given for BWR and PWR plants. The quantification of the fuel release and the tramp uranium background can be calculated with different, measurable nuclides in BWR and PWR plants. (orig.)

  4. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO2 fuel in Japan. The design concept should be compatible with UO2 fuel design. High burnup UO2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO2 8 x 8 array fuel developed for a second step of UO2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  5. Design and axial optimization of nuclear fuel for BWR reactors

    International Nuclear Information System (INIS)

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  6. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  7. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  8. Validation and application of TRACE for transient analysis of a German BWR plant

    International Nuclear Information System (INIS)

    The Karlsruhe Institute of Technology (KIT) is participating on CAMP to validate TRACE for LWR transient analysis. The validation of safety-relevant heat transfer models of TRACE, data BFBT bundle tests are available for institutions taking part on international OECD/NEA Benchmarks. In the first step, the BWR-relevant models of TRACE were validated using experimental data. Then, an integral plant model of a German BWR of Type-72 was developed using the multidimensional VESSEL-component of TRACE to simulate a plant events such as the TUSA. In this paper, the validation work, the plant modeling and selected results will be presented and discussed. (author)

  9. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    The PPD (Nuclear Power Plant Data Base) has been under development in JAERI since 1983 as a six-year program to provide useful information for reactor safety regulation and reactor safety research. Information source of the PPD is mainly based on SAR's (Safety Analysis Reports) of 35 nuclear power plants which are operating, under construction or under licensing review in Japan. The report BWR edition consists of several data lists stored in the PPD, concerning safety design such as performances, equipments and installations of 18 BWR plants in Japan. The informations are based on SAR Attachment Chapter 8 ''Nuclear Reactor Facility Safety Design''. (author)

  10. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

    First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

  11. Pool swell in a nuclear containment wetwell. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, T.

    1976-04-01

    A brief description is presented of scale model tests conducted to study LOCA induced wetwell pool swelling in the BWR Mk 1 containment pressure suppression system. The Mk 1 containment configuration is described together with the scale model design, the conduct of the tests, and the experimental results. (DG)

  12. Development of advanced BWR

    International Nuclear Information System (INIS)

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  13. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  14. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  15. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    The development of the database called PPD (Nuclear Power Plant Database) has started in 1983 at JAERI as a six-year program to provide useful information for reactor safety regulation and reactor safety research. In 1988 the program has been accomplished, and since then the data in the database has been updating and adding. Information source of the PPD is based on SAR's (Safety Analysis Report) of 47 nuclear power plants which are operating, under construction or under licensing review in Japan. The report, BWR edition, consists of lists of major data stored in the PPD, relating to safety design of 25 BWR plants in Japan. (author)

  16. Nuclear power plants

    International Nuclear Information System (INIS)

    Data concerning the existing nuclear power plants in the world are presented. The data was retrieved from the SIEN (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: nuclear plants, its status and type; installed nuclear power plants by country; nuclear power plants under construction by country; planned nuclear power plants by country; cancelled nuclear power plants by country; shut-down nuclear power plants by country. (E.G.)

  17. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR; BUTREN-RC un sistema hibrido para la optimizacion de recargas de combustible nuclear en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Castillo M, J.A. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  18. Standard Technical Specifications General Electric plants, BWR/4:Bases (Sections 3.4-3.10). Volume 3, Revision 1

    International Nuclear Information System (INIS)

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the specifications for all chapters and sections of the improved STS. Volume 2 contains he Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. This document, Volume 3, contains the Bases for Sections 3.4-3.10 of the improved STS

  19. Standard Technical Specifications General Electric plants, BWR/6: Bases (Sections 3.4-3.10). Volume 3, Revision 1

    International Nuclear Information System (INIS)

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. This document, Volume 3 contains the Bases for Sections 3.4-3.10 of the improved STS

  20. Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

    International Nuclear Information System (INIS)

    The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

  1. Results of the benchmarking in radiological protection practices during fuel reloads in the nuclear power plants of Limerick (BWR) and Ginna (PWR) in the United States of North America

    International Nuclear Information System (INIS)

    The nuclear power plant of Laguna Verde, unique in our country, has been imposed several goals related with the continuous improvement of their acting; increase in the quantity of continuous days for operation cycle, improvement in the chemical indexes of the reactor coolant, improvement in the indexes of nuclear security, improvement in the indicators of industrial security, improvement in the standards of radiological protection, etc.; in this last item is precisely where is necessary to search creative solutions to be able to maintain the collective doses of the personnel so low as reasonably it is possible (ALARA) especially due to the last projects of extension of useful life of the nuclear power plant (zinc injection, noble metals and hydrogen) and of power increment of the nuclear power plant of Laguna Verde, same that represent in the short period an increment of collective dose and of exposition levels (until 200%) in very specific points of the primary systems of the reactor. (Author)

  2. Validation of BWR advanced core and fuel nuclear designs with power reactor measurements

    International Nuclear Information System (INIS)

    Power reactor measurements have been important in validating the reliability, performance characteristics and economics of BWR advanced core and fuel designs. Such measurements go beyond the data obtainable from normal reactor operation and provide detailed benchmark data necessary to verify design and licensing computer design and simulation models. In some cases, such as in the validation of the performance of zirconium barrier pellet-cladding-interaction (PCI) resistant cladding, the BWR power reactor measurements have subjected the advanced fuel design to operating conditions more severe than normal operating conditions, thereby providing nuclear-thermal-mechanical-corrosion performance data for accelerated or extended conditions of operation. In some cases destructive measurements have been carried out on BWR power reactor fuel to provide microscopic and macroscopic data of importance in validating design and licensing analysis methods. There is not uniform agreement among core and fuel designers on the needs for special power reactor core and fuel measurements for validation of advanced designs. The General Electric approach has been to error on the side of extensive, detailed measurements so as to assure reliable performance licensing and economic design and predictive capability. This paper is a summary of some of the validative power reactor measurements that have been carried out on advanced BWR core and fuel designs. Some comparisons of predictions with the data are summarized

  3. Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

  4. BWR 90: An evolutionary ABWR plant for the next decade(s)

    International Nuclear Information System (INIS)

    The future of nuclear power is determined by ''conflicting'' considerations. It must be generated at acceptable cost with negligible environmental effects, and it must achieve and maintain a reasonable level of public acceptance. The latter is, however, not necessarily governed by rational assessments. The ABB approach to ''nuclear'' can be characterized as a ''cautious evolution''. In the next decade(s), its offerings will largely be based on ''evolutions'' of two successful LWR types, the BWR 90 and System 80+. Both designs can be designed, licensed and constructed in accordance with any safety regulations in force or envisaged in the Western world. This paper elaborates on BWR 90, a design that is closely based on its forerunner, taking into account experiences from design and engineering, construction and commissioning, and operation, needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings

  5. Operator training simulator for nuclear power plant

    International Nuclear Information System (INIS)

    In nuclear power plants, training of the operators is important. In Japan, presently there are two training centers, one is BWR operation training center at Okuma-cho, Fukushima Prefecture, and another the nuclear power generation training center in Tsuruga City, Fukui Prefecture, where the operators of PWR nuclear power plants are trained. This report describes the BWR operation training center briefly. Operation of a nuclear power plant is divided into three stages of start-up, steady state operation, and shut down. Start-up is divided into the cold-state start-up after the shut down for prolonged period due to periodical inspection or others and the hot-state start-up from stand-by condition after the shut down for a short time. In the cold-state start-up, the correction of reactivity change and the heating-up control to avoid excessive thermal stress to the primary system components are important. The BWR operation training center offers the next three courses, namely beginner's course, retraining course and specific training course. The training period is 12 weeks and the number of trainees is eight/course in the beginner's course. The simulator was manufactured by modeling No. 3 plant of Fukushima First Nuclear Power Station, Tokyo Electric Power Co. The simulator is composed of the mimic central control panel and the digital computer. The software system comprises the monitor to supervise the whole program execution, the logic model simulating the plant interlock system and the dynamic model simulating the plant physical phenomena. (Wakatsuki, Y.)

  6. Feasibility study on development of plate-type heat exchanger for BWR plants

    International Nuclear Information System (INIS)

    In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

  7. On-line critical control rod pattern prediction algorithm for BWR plant startup

    International Nuclear Information System (INIS)

    This paper describes an on-line algorithm for predicting the critical control rod pattern, which has been developed to reduce the mental strain on operators while withdrawing control rods in the BWR plant startup operation. The proposed algorithm estimates a target eigenvalue (eigenvalue bias) for a three-dimensional neutron kinetics model with a neutron source incorporating actual neutron detector readings. The critical control rod pattern is then predicted based on the estimated eigenvalue bias. The algorithm has been verified using data obtained from an actual startup operation on a BWR model-5 plant, and the estimated eigenvalue bias agreed well with the effective multiplication factor at the criticality actually determined from the operator's judgement. (author)

  8. Natural heat transfer augmentation in passive advanced BWR plants

    International Nuclear Information System (INIS)

    In the European Simplified Boiling Water Reactor (ESBWR), the long-term post-accident containment pressure is determined by the combination of non condensable gas pressure and steam pressure in the wet well gas space. Since there are no active systems for heat removal in the wet well, energy transmitted to the wet well gas space, by a variety of means, must be removed by passive heat transfer to the walls and suppression pool (SP). The cold suppression pool located below the hotter gas space provides a stable configuration in which convection currents are suppressed thus limiting heat and mass transfer between the gas space and pool. However, heat transfer to the walls results in natural circulation currents that can augment the heat and mass transfer to the pool surface. Using a simplified model, parametric studies are carried out to show that augmentation of the order of magnitude expected can significantly impact the heat and mass transfer to the pool. Additionally a review of available literature in the area of augmentation and mixed convection of this type is presented and indicates the need for additional experimental work in order to develop adequate models for heat and mass transfer augmentation in the configuration of a BWR suppression pool. (author)

  9. Development of remote automatic equipment for BWR power plants

    International Nuclear Information System (INIS)

    The development of remote control, automatic equipment for nuclear power stations has been promoted to raise the rate of operation of plants by shortening regular inspection period, to improve the safety and reliability of inspection and maintenance works by mechanization, to reduce the radiation exposure dose of workers and to reduce the manpower required for works. The taking-off of control rod drives from reactors and fixing again have been mechanized, but the disassembling, cleaning, inspection and assembling of control rod drives are manually carried out. Therefore, Hitachi Ltd. has exerted effort to develop the automatic equipment for this purpose. The target of development, investigation, the construction and function of the equipment, the performance and the effect of adopting it are reported. The equipment for the volume reduction of spent fuel channel boxes and spent control rods is developed since these are major high level radioactive solid wastes, and their apparent volume is large. Also the target of development, investigated things, the construction and function of the equipment, the performance and the effect of adopting it are reported. (Kako, I.)

  10. Nuclear power plant control system

    International Nuclear Information System (INIS)

    Purpose: To effectively transfer the operation of a nuclear power plant including a bwr type reactor to single load operation in the station with no increase in the neutron flux and water level in the reactor by the combined use of recycle pump trip and feedwater pump trip. Method: Upon rapid load decrease in a turbine generator, at least one of usually operated feedwater pumps and a recycle pump are tripped and the starting for a stand-by feedwater pump is inhibited. This rapidly decreases the recycling flow rate to thereby lower the neutron flux before generation of pressure increase and also decreases the feedwater flow rate to thereby suppress the increase in the water level due to increase in voids, whereby the operation is smoothly transferred to the single load operation in the station. (Horiuchi, T.)

  11. Energy investment in nuclear power plants

    International Nuclear Information System (INIS)

    The energy investment in terms of electrical and thermal energy requirements for nuclear power plants is examined. The total lifetime energy inputs required for a 1000-MW(e) plant based on a 30-year plant lifetime and 0.75 plant factor are presented for several pressurized-water-reactor (PWR) and boiling-water-reactor (BWR) systems, two high-temperature gas-cooled reactor (HTGR) systems, and one heavy-water-reactor (HWR) system. The energy analyses reviewed here have demonstrated that the energy expenditures for the various reactor systems are very much less than the energy produced by the reactors and that the initial energy investments are recovered in a short time after startup--several months to 2 years. The energy requirements associated with the individual processes, such as mining, enrichment, construction, and waste disposal, are tabulated for two different fuel cycles for a PWR and a BWR. The enriching process is by far the largest component of the electrical requirements, representing 85 to 90% of the total electrical energy investment. The energy used in constructing and operating the reactor constitutes the largest single investment of thermal energy, representing 50% of the total thermal energy requirements. Results of several analyses are examined and comparisons made between nuclear power plants, a coal plant, and a solar thermal-conversion plant

  12. Construction and operation of nuclear power plants

    International Nuclear Information System (INIS)

    How does a nuclear power plant work. Which reactor types are in use. What safety measures are being taken. These questions and the like are frequently asked by those interested in nuclear power generation. The respective answers are to be found in the report ''Construction and Operation of Nuclear Power Plants''. Nuclear-physical fundamentals and the basic safety measures are explained, and four reactor types that are most common in the Federal Republic of Germany are described: PWR-, BWR-, HTR-type and faster breeder reactors. For each reactor type, the principle of operation, steam generator system, auxiliary and service buildings as well as the respective safety devices are indicated, and visualized by means of numerous illustrations. The report is meant to be instrumental to the purpose of getting objectiveness into the public discussion on the peaceful use of nuclear energy. (orig.)

  13. Nuclear plant

    International Nuclear Information System (INIS)

    The reception area of a spent fuel centre (back end plant) contains a decontamination device for fuel element transport containers. A two part, symmetrical working platform is used to monitor their surfaces. Both parts can be pushed out of the way and can be adjusted in heighth. A spindle motor and hydraulic motors are used for this purpose. (DG)

  14. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    International Nuclear Information System (INIS)

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  15. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  16. Development of a nuclear pump seal for BWR application

    International Nuclear Information System (INIS)

    A new, high integrity seal was developed for reactor recirculation pumps at the Nine Mile Point Unit 1 (NMP-1) plant. The target was to extend seal lifetime to four years and to fit with minimum modification within the original seal cartridge envelope. Laboratory simulation tests were conducted on the original, then the new seals. Responses to the most trying conditions were compared, i.e., extended low pressure running, hot standby, transients of pressure, temperature, speed and shaft movement, and overpressure on one seal stage. These conditions revealed periods of insufficient leakage flow (lubrication) and vaporization between the original seal faces, causing wear and instability. The new seal was designed to avoid this. It was thoroughly proven as a single seal, followed by 6,000 h of accelerated testing of a complete cartridge. Cartridges of the new seals were first installed in two of the five pumps at NMP-1 in the spring of 1986. After one fueling cycle of trouble-free service, one was examined two years later and found in excellent condition. Its sister cartridge was therefore left in service for another cycle, with seals of the new design simultaneously being installed in the remaining three recirculation pumps. From 1988 to mid 1990, the new seals saw standstill, flooded conditions while the plant went through a general upgrade program. It was decided not to examine or refurbish them prior to plant restart, but they have since continued to perform perfectly. One cartridge of the new seal design was installed in the Oyster Creek plant in March 1990. This was observed during 1990 to give particularly stable performance (staging pressure) compared with remaining seals of the original design. Other new seals were therefore installed during the spring, 1991 fueling outage

  17. Revaluation on measured burnup values of fuel assemblies by post-irradiation experiments at BWR plants

    International Nuclear Information System (INIS)

    Fuel composition data for 8x8 UO2, Tsuruga MOX and 9x9-A type UO2 fuel assemblies irradiated in BWR plants were measured. Burnup values for measured fuels based on Nd-148 method were revaluated. In this report, Nd-148 fission yield and energy per fission obtained by burnup analyses for measured fuels were applied and fuel composition data for the measured fuel assemblies were revised. Furthermore, the adequacies of revaluated burnup values were verified through the comparison with burnup values calculated by the burnup analyses for the measured fuel assemblies. (author)

  18. HVDC transmission from nuclear power plant

    International Nuclear Information System (INIS)

    The HVDC transmission directly from nuclear power plants is one of the patterns of long distance and large capacity HVDC transmission systems. In this report, the double pole, two-circuit HVDC transmission from a BWR nuclear power plant is considered, and the dynamic response characteristics due to the faults in dc line and ac line of inverter side are analyzed, to clarify the dynamic characteristics of the BWR nuclear power plant and dc system due to system faults and the effects of dc power control to prevent reactor scram. (1) In the instantaneous earthing fault of one dc line, the reactor is not scrammed by start-up within 0.8 sec. (2) When the earthing fault continues, power transmission drops to 75% by suspending the faulty pole, and the reactor is scrammed. (3) In the instantaneous ground fault of 2 dc lines, the reactor is not scrammed if the faulty dc lines are started up within 0.4 sec. (4) In the existing control of dc lines, the reactor is scrammed when the ac voltage at an ac-dc connection point largely drops due to ac failure. (J.P.N.)

  19. Investigation of control rod worth and nuclear end of life of BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of {sup 10}B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% {sup 10}B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in {sup 10}B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming.

  20. Investigation of control rod worth and nuclear end of life of BWR control rods

    International Nuclear Information System (INIS)

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

  1. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  2. Real time simulation of the main steam system of a BWR nuclear power station

    International Nuclear Information System (INIS)

    This paper presents a real time model of the main steam system for a BWR 675 MW power plant unit. The model includes the start up and shut down of the system, where the steam flow is very small or non existent and phenomena like condensation can occur, changing drastically the effects observed from those of normal operation at medium or high loads. Severe transients are also contemplated. Consistency and stability tests were done to the model, and it was validated for steady state using plant design data. During transients the model's results were compared with the predictions of the Final Safety Analysis Report (FSAR) for the prototype unit, and it was found that the model's response follow the expected trends

  3. Development of large-capacity main steam isolation valves and safety relief valves for next-generation BWR plant

    International Nuclear Information System (INIS)

    A study was made of high capacity main steam isolation valves (MSIV) and safety relief valves (SRV) for the main steam line of a boiling water reactor (BWR). The next-generation BWR plants, which are planned to have higher thermal power, have raised concerns relating to the main steam line of an increase in maintenance work to SRVs and erosion of the MSIV valve seat due to the increased main steam flow velocity. In this research project, the capacity of the MSIV and SRV was increased and the valve configuration was changed in an attempt to solve these problems. (authors)

  4. Study on the feasibility of 1300 MWe class simplified BWR plant

    International Nuclear Information System (INIS)

    A range of power levels for 1000 MWe-1500 MWe natural circulation core was found to be feasible from the thermal hydraulic performance standpoint by our sensitivity analysis. In this study, we selected a power level of 1300 MWe that is expected to satisfy Japanese Utilities needs. After we set the RPV configuration, we will study the detailed comprehensive analysis so that we can confirm the technical feasibility of large scaled simplified BWR. RPV inner diameter 7.5 m, which can be manufactured with current technology and present facilities, and the chimney height of 8.5 m was selected. After a preliminary design of the core and fuel was carried out, the natural circulation core flow was calculated by EASHAP code. The stability evaluation during normal operation is analyzed and a major transient analysis is conducted. The design of the core and fuel is evaluated based on PANACEA code. The detailed analysis shows that a 1300 MWe class natural circulation core satisfies the thermal and stability criteria. The containment system, which consists of the drywell and suppression chamber, is determined with supporting containment pressure-temperature analytical response. The layout inside the primary containment vessel that is applicable to a RPV incorporating the 1300 MWe core is approximately arranged. From the above, it is confirmed that 1000 MWe is not technical upper power limit of the simplified BWR plant. (author)

  5. Optimization of individual monitoring for β radiation dose at BWR plants in Japan

    International Nuclear Information System (INIS)

    β as well as γ-ray individual monitoring is carried out for all radiation workers at BWR plants in Japan. β-ray individual exposure has hardly been found so far. A series of β-ray distribution survey were performed at three BWR plants to optimize individual monitoring for β-rays. β and γ ray dose-equivalent rates were measured using a thin-window type ionization chamber survey meter equipped with an acrylic board for the β-ray shield, for equipment and instrument installed at reactor primary cooling system, main steam system, condensate and feedwater system, and off-gas system which are required to be inspected periodically. Residual maximum energy for β particles was calculated after β-ray absorption curve was made, based on ISO 6980, from measured data. Shield effect of working clothes which had been worn at the inspection was confirmed. a. β/γ ray dose-equivalent ratio and residual maximum energy for β particles: The β/γ ratio fairly exceeded ten times and the residual maximum energy for β particles was more than 1 MeV from the results monitored at the off-gas system. On the other hand, the β/γ ratio was less than one at most of the other systems and the residual maximum energy was less than 0.5 MeV at all of the systems except the off-gas system. b. Shield effect results of the working clothes: The working clothes protected more than 70-80% of the β-ray at the place where the residual maximum energy was less than 0.5 MeV. However, about 10-30% of the β ray was only protected by the clothes at the off - gas system where the residual maximum energy was high. It is suggested that dose equivalent for tissues of radiation workers can be controlled by individual monitoring for γ radiation dose only for most of the systems at BWR plants because the β/γ ratio is less than one, the residual maximum energy of beta particles is less than 0.5 MeV, and working clothes are effective in shielding workers from the radiation. As for the off-gas system

  6. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

    During the next decade a rise in the energy demand is expected worldwide, and this will in particular call for electricity generation capacity. A number of old generating plants, both nuclear and other plants, will probably have to be shut down for aging reasons, and their replacement will enhance the need for new generating capacity. The ABB Atom considers this situation to be met with a 'cautious evolution'. The offerings will largely be based on 'evolutions' of the successful light water reactor BWR 75. The new, evolutionary plant design of ABB Atom is the BWR 90. It can be designed, licensed and constructed in accordance with any safety regulations now in force or envisaged in the Western world. Emphasis has been, and will be, placed on features that facilitates licensing, shortens construction time and keeps electricity generation costs favourable. ABB also continues to develop a design of the 'passive' type, such as the 'passive' PIUS system, for possible deployment in the future. These efforts are more long-term activities, since development, verification and licensing of distinctly 'new' reactor concepts will have an extensive lead time. This paper presents the BWR 90 and its current status. The design is based on that of its forerunner, the BWR 75 standard design, taking into account the experiences gained from design and engineering, construction, commissioning, and operation of BWR 75 plants, the needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings. (author) 4 figs

  7. On-line BWR materials monitoring and plant component lifetime prediction

    International Nuclear Information System (INIS)

    Environmentally-assisted cracking (e.g., stress corrosion, corrosion fatigue) is a potential life-limiting phenomenon for Light Water Reactors and has been observed in piping, pressure vessels, core-internals, steam generators, etc. A methodology for predicting a component's life-time and for recommending actions for life extension is presented which is based on: monitoring the extent of cracking damage in selected reference or key components; establishing the system's environmental and material conditions; and predicting, via fundamental understanding of the cracking process, the future behavior of all components for defined plant conditions. A life prediction and extension methodology which combines all of these elements may be applied to piping, core and pressure vessel components; in this article, examples are given of its application to BWR piping

  8. Assessment of hydrogen combustion effects in the BWR/6 - Mark III Standard Plant

    International Nuclear Information System (INIS)

    This report discusses General Electric's study of potential hydrogen combustion effects on the Standard Mark III containment during postulated severe accidents. This study was performed as part of the Probabilistic Risk Assessment of the BWR/6 - Mark III Standard Plant. The methodology of determining the accident event sequence and modeling of the Boiling Water Reactor core response, including hydrogen generation by metal-water reaction, is described. Combustion of hydrogen released to the containment is analyzed and effects on the Mark III containment system are assessed. It is concluded that even for those cases where containment integrity may be lost, the containment function (i.e., limiting offsite doses) is maintained by the drywell and suppression pool

  9. Nuclear power plant outages

    International Nuclear Information System (INIS)

    The Finnish Radiation and Nuclear Safety Authority (STUK) controls nuclear power plant safety in Finland. In addition to controlling the design, construction and operation of nuclear power plants, STUK also controls refuelling and repair outages at the plants. According to section 9 of the Nuclear Energy Act (990/87), it shall be the licence-holder's obligation to ensure the safety of the use of nuclear energy. Requirements applicable to the licence-holder as regards the assurance of outage safety are presented in this guide. STUK's regulatory control activities pertaining to outages are also described

  10. Status report on the application of process noise technique in nuclear power plants

    International Nuclear Information System (INIS)

    The report gives a survey of applications of noise technique reported for nuclear power plants. The scope has been limited to areas of interest for BWR and PWR plants of the types found in Sweden and with an emphasis on cases where the practical applicability has been clearly demonstrated. (author)

  11. Improvement of fire protection measures for nuclear power plants

    International Nuclear Information System (INIS)

    Improvements of fire protection measures for nuclear power plants were performed as following items: Collection of necessary data and evaluation of parameters, Walk down in the typical BWR and PWR plants, Application of developed analysis method to actual plants, Development of fire hazard analysis manual. With regard to fire tests for the fire data acquisition, the test plan was made about cable fire, oil fire and electric cabinet fire in the concrete test facilities. Implementation of fire hazard analysis model and simulation were performed as following items: Fire analysis codes FDS, CFAST were implemented in order to analyze the fire progression phenomena, Trial calculation of fire hazard for typical NPP by FDS were performed for the Auxiliary feed water pump room (PWR) and the High pressure coolant flooder pump room (BWR). (author)

  12. Nuclear Power Plants. Revised.

    Science.gov (United States)

    Lyerly, Ray L.; Mitchell, Walter, III

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

  13. Training of nuclear power plant personnel practices in Finland

    International Nuclear Information System (INIS)

    Four nuclear power units are in commercial operation in Finland. TVO operates two 735 MW BWR units of Swedish design and IVO two 465 PWR units of Soviet design. Good availability figures have been reached during the last five years for all plants. Many different factors have been contributing to the success of nuclear power in Finland. The advanced training of the plant personnel combined with a high educational level of plant personnel is undoubtedly one of the most important factors. Another important factor has been the low turnover rate of the staff at the plant and in the nuclear industry in general. The favorable regulatory climate in Finland with open and trustworthy relations between the power companies and the authorities has also been an important contribution to the success. The paper gives an overview of the training of nuclear power plant personnel as seen from authorities', the power companies' and the research institutions' points of view

  14. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

    Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

  15. Safety Evalution Report related to the final design approval of the GESSAR II BWR/6 nuclear island design (Docket No. 50-447)

    International Nuclear Information System (INIS)

    This report supplements the GESSAR II SER (NUREG-0979), issued in April 1983, summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design. The review is carried out in accordance with the procedures for demonstrating the acceptability of the design for the severe-accident concerns described in draft NUREG-1070, NRC Policy on Future Reactor Designs: Decisions on Severe Accident Issues in Nuclear Power Plant Regulation. Supplement 2 also provides more recent information regarding resolution or update of the confirmatory items and FDA-1 conditions identified in SSER 1. Subject to favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in draft NUREG-1070

  16. Nuclear power plant construction

    International Nuclear Information System (INIS)

    The legal aspects of nuclear power plant construction in Brazil, derived from governamental political guidelines, are presented. Their evolution, as a consequence of tecnology development is related. (A.L.S.L.)

  17. Nuclear power plant

    International Nuclear Information System (INIS)

    In a condensate cleanup system and a reactor water cleanup system of a BWR-type reactor, in which primary coolants flow, there is disposed a filtering and desalting device using hollow thread membrane filter and ion exchange resin for a condensate cleanup system, and using a high temperature filter made of a metal, a metal oxide or ceramics as a filtering material and a precoat filter made of a powdery ion exchange resin as a filtering material for a reactor water cleanup system. This can completely remove cruds generated in the condensate system. Since the reactor water cleanup system comprises the powdery resin precoat-type filtering and desalting device and the high temperature filter using ceramics, ionic impurities such as radioactive materials can be removed. Accordingly, cruds are not carried into the inside of the reactor, and since the radioactive concentration in the reactor water is reduced, radiation exposure upon periodical inspection can be minimized almost to zero, to attain a clean plant. (T.M.)

  18. Nuclear Power Plant Technician

    Science.gov (United States)

    Randall, George A.

    1975-01-01

    The author recognizes a body of basic knowledge in nuclear power plant technoogy that can be taught in school programs, and lists the various courses, aiming to fill the anticipated need for nuclear-trained manpower--persons holding an associate degree in engineering technology. (Author/BP)

  19. Nuclear power plant erection

    International Nuclear Information System (INIS)

    The erection of a nuclear power plant covers all the installation operations related to mechanical and electrical equipment in buildings designed for this specific purpose. Some of these operations are described: erection of the nuclear boiler, erection work carried out in the building accomodating the nuclear auxiliary and ancillary equipment and the methods and the organization set up in order to carry out this work satisfactorily are analyzed

  20. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    International Nuclear Information System (INIS)

    This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

  1. Serpent: an alternative for the nuclear fuel cells analysis of a BWR; SERPENT: una alternativa para el analisis de celdas de combustible nuclear de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silva A, L.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: lidi.s.albarran@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In the last ten years the diverse research groups in nuclear engineering of the Universidad Nacional Autonoma de Mexico and Instituto Politecnico Nacional (UNAM, IPN), as of research (Instituto Nacional de Investigaciones Nucleares, ININ) as well as the personnel of the Nuclear Plant Management of the Comision Federal de Electricidad have been using the codes Helios and /or CASMO-4 in the generation of cross sections (X S) of nuclear fuel cells of the cores corresponding to the Units 1 and 2 of the nuclear power plant of Laguna Verde. Both codes belong to the Studsvik-Scandpower Company who receives the payment for the use and their respective maintenance. In recent years, the code Serpent appears among the nuclear community distributed by the OECD/Nea which does not has cost neither in its use neither in its maintenance. The code is based on the Monte Carlo method and makes use of the processing in parallel. In the Escuela Superior de Fisica y Matematicas of the IPN, the personnel has accumulated certain experience in the use of Serpent under the direction of personal of the ININ; of this experience have been obtained for diverse fuel burned, the infinite multiplication factor for three cells of nuclear fuel, without control bar and with control bar for a known thermodynamic state fixed by: a) the fuel temperature (T{sub f}), b) the moderator temperature (T{sub m}) and c) the vacuums fraction (α). Although was not realized any comparison with the X S that the codes Helios and CASMO-4 generate, the results obtained for the infinite multiplication factor show the prospective tendencies with regard to the fuel burned so much in the case in that is not present the control bar like when it is. The results are encouraging and motivate to the study group to continue with the X S generation of a core in order to build the respective library of nuclear data as a following step and this can be used for the codes PARCS, of USA NRC, DYN3D of HZDR, or others developed locally

  2. Regression analysis of nuclear plant capacity factors

    International Nuclear Information System (INIS)

    Operating data on all commercial nuclear power plants of the PWR, HWR, BWR and GCR types in the Western World are analysed statistically to determine whether the explanatory variables size, year of operation, vintage and reactor supplier are significant in accounting for the variation in capacity factor. The results are compared with a number of previous studies which analysed only United States reactors. The possibility of specification errors affecting the results is also examined. Although, in general, the variables considered are statistically significant, they explain only a small portion of the variation in the capacity factor. The equations thus obtained should certainly not be used to predict the lifetime performance of future large reactors

  3. Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

    2011-07-01

    Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

  4. New nuclear plant design and licensing process

    International Nuclear Information System (INIS)

    This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

  5. Kruemmel nuclear power plant

    International Nuclear Information System (INIS)

    This short description of the site and the nuclear power plant with information on the presumable effects on the environment and the general public is to provide some data material to the population in a popular form so that the citizens may in form themselves about the plant. In this description which shall be presented to the safety report, the site, the technical design and the operation mode of the nuclear power plant are described. Some problems of the emission and the effects of radioactive materials as well as other issues related to the plant which are of interest to the public are dealt with. The supposed accidents and their handling are discussed. The description shows that the selected site is suitable for both setting-up and operation of the plant without affecting the safety of the people living there and that in admissible burdens of the environment shall not have to be expected. (orig./HP)

  6. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

    International Nuclear Information System (INIS)

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

  7. Nuclear power plant simulator

    International Nuclear Information System (INIS)

    In this paper, real time nuclear power plant simulator for student education is described. The simulator is composed of a hybrid computer and an operating console. Simulated power plant is a 36 MWt PWR plant, and the average temperature of the primary coolant within the reactor is controlled to be constant. Reactor Kinetics, fuel temperature, primary coolant temperature, temperature and pressure of steam within the steam generator, steam flow, control rod driving system, and feed water controlling system are simulated. The use of the hybrid computer made it possible to simulate a relatively large scale power plant with a comparatively small size computing system. (auth.)

  8. Specimen sampling system in a BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Purpose: To accurately determine the impurity concentration in primary coolants. Constitution: Gold or platinum coating is applied to the inner surface of pipeways for a condensate specimen sampling system led out between a condensator and a low pressure condensate pump, a feed water specimen sampling system branched from a feed water pipeway at the downstream of a feed water heater and a reactor water specimen sampling system connected to a reactor water clean up line lead out from a reactor pressure vessel. Since these pipeways have inner diameter of about 6 mm and length of about 100 mm, stainless steel pipeways applied with platings to the inner surface are preferably used in view of the strength. Since gold or platinum is very much stable chemically at high temperature or low temperature, it neither adsorbs metal ions in the specimen water nor erode to leach out impurities, whereby errors due to the pipeway can be eliminated. (Sekiya, K.)

  9. Planning guidance for nuclear-power-plant decontamination. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munson, L.F.; Divine, J.R.; Martin, J.B.

    1983-06-01

    Direct and indirect costs of decontamination are considered in the benefit-cost analysis. A generic form of the benefit-cost ratio is evaluated in monetary and nonmonetary terms, and values of dollar per man-rem are cited. Federal and state agencies that may have jurisiction over various aspects of decontamination and waste disposal activities are identified. Methods of decontamination, their general effectiveness, and the advantages and disadvantages of each are outlined. Dilute or concentrated chemical solutions are usually used in-situ to dissolve the contamination layer and a thin layer of the underlying substrate. Electrochemical techniques are generally limited to components but show high decontamination effectiveness with uniform corrosion. Mechanical agents are particularly appropriate for certain out-of-system surfaces and disassembled parts. These processes are catagorized and specific concerns are discussed. The treatment, storage, and disposal or discharge or discharge of liquid, gaseous, and solid wastes generated during the decontamination process are discussed. Radioactive and other hazardous chemical wastes are considered. The monitoring, treatment, and control of radioactive and nonradioactive effluents, from both routine operations and possible accidents, are discussed. Protecting the health and safety of personnel onsite during decontamination is of prime importance and should be considered in each facet of the decontamination process. The radiation protection philosophy of reducing exposure to levels as low as reasonably achievable should be stressed. These issues are discussed.

  10. Recent developments in BWR water chemistry

    International Nuclear Information System (INIS)

    Water chemistry is of critical importance to the operation and economic viability of the Boiling Water Reactor (BWR). A successful water chemistry program will satisfy the following goals: - Minimize the incidence and growth of SCC/IASCC, - Minimize plant radiation fields controllable by chemistry, -Maintain fuel integrity by minimizing cladding corrosion, - Minimize flow-accelerated corrosion (FAC) in balance-of-plant components. The impact of water chemistry on each of these goals is discussed in more detail in this paper. It should be noted that water chemistry programs also include surveillance and operating limits for other plant water systems (e.g., service water, closed cooling water systems, etc.) but these are out of the scope of this paper. This paper reviews developments in water chemistry guidelines for U.S. BWR nuclear power plants. (author). 2 figs., 2 tabs., 7 refs

  11. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    International Nuclear Information System (INIS)

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

  12. Understanding the causes of past nuclear plant availability performance through a multiple regression technique

    International Nuclear Information System (INIS)

    An analysis of past nuclear power plant availability performance is presented which covers the experience of 72 U.S. BWR's and PWR's currently in operation. This analysis quantitatively related availability to several design and organizational characteristics, including: plant size, age, staffing levels, maintenance quality, turnover rates, and other factors. The results are presented in terms of Physical (design), organizational, and external factors affecting plant performance

  13. Nuclear power plants

    International Nuclear Information System (INIS)

    Before the economical adaptability of nuclear power plants was achieved, many ways were tried to technically use nuclear fission. In the course of a selection process, of numerous types of reactors, only a few have remained which are now taking part in the competition. The most important physical fundamentals, the occurence of various reactor concepts and the most important reactor types are the explained. (orig./TK)

  14. Long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Considering next generation requirement for nuclear plants, a long cycle operating simplified BWR (LSBWR) concept is proposed. The major features of LSBWR are; 1) Long cycle operation core using uranium fuels; 2) Simplified system and component as well as passive systems; 3) Combined building concept with ship hull structure. This concept have potential to reduce construction cost and to Increase availability. Safety feature of LSBWR makes possible to attain no evacuation capability in case of a severe accident. Further research and development is underway. (author)

  15. Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code

    International Nuclear Information System (INIS)

    The Monte Carlo N-Particle transport code (MCNP) was adopted to calculate the skyshine dose from the turbine building of a BWR plant for obtaining precise estimations at the site boundary. In MCNP calculation, the equipment and piping arranged on the operating floor of the turbine building were considered and modeled in detail. The inner and outer walls of the turbine building, the shielding materials around the high-pressure turbine, and the piping connected from the moisture separator to the low-pressure turbine were all considered. A three-step study was conducted to estimate the applicability of MCNP code. The first step is confirming the propriety of calculation models. The atmospheric relief diaphragms, which are installed on top of the low-pressure turbine exhaust hood, are not considered in the calculation model. There was little difference between the skyshine dose distributions that were considered when using and not using the atmospheric relief diaphragms. The calculated dose rates agreed well with the measurements taken around the turbine. The second step is estimating the dose rates on the outer roof surface of the turbine building. This calculation was made to confirm the dose distribution of gamma-rays on the turbine roof before being scattered into the air. The calculated dose rates agreed well with the measured data. The third step is making a final confirmation by comparing the calculations and measurements of skyshine dose rates around the turbine building. The source terms of the main steam system are based on the measured activity data of N-16 and C-15. As a conclusion, we were able to calculate reasonable skyshine dose rates by using MCNP code. (authors)

  16. BWR 90 - the advanced BWR of the 1990s

    International Nuclear Information System (INIS)

    The future global role of nuclear power will be determined by its ability to provide economical and safe energy. Nuclear power, like any other substantial contributor to the world's energy needs, must be generated at an acceptable cost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a 'cautious evolution'; for the next decade the company will largely base its offerings to the market on its 'evolutionary' light water reactor design, the BWR 90. This design builds closely on the experience from successful construction and operation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average load factor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very best performing plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR design development focuses on meeting requirements from utilities as well as new regulatory requirements. A particular emphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimization of buildings and containment design to decrease construction time and costs, and selection of materials as well as maintenance and operating procedures to even further reduce occupational radiation exposures. (orig.)

  17. Analysis of BWR OPRM plant data and detection algorithms with DSSPP

    International Nuclear Information System (INIS)

    All U.S. BWRs are required to have licensed stability solutions that satisfy General Design Criteria (GDC) 10 and 12 of 10 CFR 50 Appendix A. Implemented solutions are either detect and suppress or preventive in nature. Detection and suppression of power oscillations is accomplished by specialized hardware and software such as the Oscillation Power Range Monitor (OPRM) utilized in Option III and Detect and Suppress Solution - Confirmation Density (DSS-CD) stability Long-Term Solutions (LTSs). The detection algorithms are designed to recognize a Thermal-Hydraulic Instability (THI) event and initiate control rod insertion before the power oscillations increase much higher above the noise level that may threaten the fuel integrity. Option III is the most widely used long-term stability solution in the US and has more than 200 reactor years of operational history. DSS-CD represents an evolutionary step from the stability LTS Option III and its licensed domain envelopes the Maximum Extended Load Line Limit Analysis Plus (MELLLA +) domain. In order to enhance the capability to investigate the sensitivity of key parameters of stability detection algorithms, GEH has developed a new engineering analysis code, namely DSSPP (Detect and Suppress Solution Post Processor), which is introduced in this paper. The DSSPP analysis tool represents a major advancement in the method for diagnosing the design of stability detection algorithms that enables designers to perform parametric studies of the key parameters relevant for THI events and to fine tune these system parameters such that a potential spurious scram might be avoided. Demonstrations of DSSPPs application are also presented in this paper utilizing actual plant THI data. A BWR/6 plant had a plant transient that included unplanned recirculation pump transfer from fast to slow speed resulting in about 100% to ∼40% rated power decrease and about 99% to ∼30% rated core flow decrease. As the feedwater temperature is reduced

  18. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  19. Nuclear plant scram reduction

    International Nuclear Information System (INIS)

    The Nuclear Utility Management and Human Resources Committee (NUMARC) is a confederation of all 55 utilities with nuclear plants either in operation or under construction. NUMARC was formed in April 1984 by senior nuclear executives with hundreds of man-years of plant experience to improve (plant) performance and resolve NRC concerns. NUMARC has adopted 10 commitments in the areas of management, training, staffing and performance. One of these commitments is to strive to reduce automatic trips to 3 per year per unit for calendar year 1985 for plants in commercial operation greater than 3 years (with greater than 25% capacity factor). This goal applies to any unplanned automatic protection system trips at any time when the reactor is critical. Each utility has committed to develop methods to thoroughly evaluate all unplanned automatic trips to identify the root causes and formulate plans to correct the root causes thus reducing future unplanned scrams. As part of this program, the Institute of Nuclear Power Operations (INPO) collects and evaluates information on automatic reactor trips. It publishes the results of these evaluations to aid the industry to identify root causes and corrective actions

  20. Initiative against nuclear power plants

    International Nuclear Information System (INIS)

    This publication of the Initiative of Austrian Nuclear Power Plant Opponents contains articles on radiactive waste dispoasal in Austria and and discusses safety issues of the nuclear power plant 'Zwentendorf'. (kancsar)

  1. Effect of fraction of voids in the nuclear fuel burned for a 10 X 10 assembly of a BWR

    International Nuclear Information System (INIS)

    A major source of uncertainty in BWR reactor physics is associated with the properties of moderation and coolant bypass regions with a very significant impact on nuclear parameters such as: the finite multiplication factor (k∞), area migration of neutrons (M2) and the void coefficient of reactivity (aν). In this work, we assess the effect caused by the presence of voids in the moderator during the burning of fuel in a fuel assembly type SVEA-96 for a BWR; the codes uses as a tool were INTERPIN-3 and CASMO-4. The geometry SVEA-96 is characterized by an assembly subdivided in four sub-bundles, through an internal bypass cross-shaped gap that allows a more uniform distribution of the moderator, providing a better distribution in the neutrons flux, and thus provide a better distribution of energy and burned. This study was conducted for a wide range of void fractions, from 0% (pure liquid) to 100% (pure steam) and covered: 1) The effect caused by the presence of voids during the burning of nuclear fuel 2) the effects of the structure of energy groups including libraries of cross sections based on ENDF/B-4, and 3) the impact of the presence of control rod. The burning range is from 0 G Wd/Mt to 50 G Wd/Mt. (Author)

  2. Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    The Nuclear Plant Data Bank (NPDB) is being developed at the Los Alamos National Laboratory to assist analysts in the rapid and accurate creation of input decks for reactor transient analysis. The NPDB will reduce the time and cost of the creation or modification of a typical input deck. This data bank will be an invaluable tool in the timely investigation of recent and ongoing nuclear reactor safety analysis. This paper discusses the status and plans for the NPDB development and describes its anticipated structure and capabilities

  3. Operating experience from Swedish nuclear power plants 2008

    International Nuclear Information System (INIS)

    The overall safety level of the Swedish nuclear power plants during 2008 has been high, despite disturbances that caused longer shutdowns in some plants. A number of improvements have been made in order further to improve the safety of the plants. Electricity production from the power plants amounted to 61.3 TWh, which represents 43 % of Sweden's total electricity production during the year. The total average energy availability was lower than in previous years. For the BWR reactors, it amounted to 78 %, as against an average value of about 84 % for the last five years. For the PWRs, the energy availability was about 87 % in 2008, which is about the same as the five-year average. Oskarshamn 3 and Forsmark 3 suffered from long shutdowns due to cracks in the reactor control rods, which was a contributory reason for the poor performance figures for the BWR reactors. On the PWR side, it was Ringhals 2's problems with auxiliary feed water system capacity that caused a longer shutdown. During the year, KSU started its courses for the maintenance staff of all Swedish power plants at the shut-down Barsebaeck site. This provides training for electrical and mechanical maintenance personnel under authentic conditions, working on real components and systems before doing the same under live conditions at their home power plants

  4. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

    Cofrentes Nuclear Power Plant (GE BWR/6) has implemented its specific Severe Accident Management Program within this year 2000. New organization and guides have been developed to successfully undertake the management of a severe accident. In particular, the Technical Support Center will count on a new ''Severe Accident Management Team'' (SAMT) which will be in charge of the Severe Accident Guides (SAG) when Control Room Crew reaches the Emergency Operation Procedures (EOP) step that requires containment flooding. Specific tools and training have also been developed to help the SAMT to mitigate the accident. (author)

  5. Robots cut risks and costs in nuclear power plants

    International Nuclear Information System (INIS)

    The electric utility industry, the third largest user of computers in the United States, is realizing the value of a relatively new computer-related technology; robotics. Since its use in the cleanup of Three Mile Island, 44 utility companies have used robotic devices to save radiation exposure to their employees and to achieve measurable and often significant cost savings. This paper reports on the use of robots to reduce risk and cost in electric utilities especially in PWR and BWR nuclear power plants

  6. Nuclear turbine power plant

    International Nuclear Information System (INIS)

    Purpose : To improve the heat cycle balance in a nuclear turbine power plant or the like equipped with a moisture separating and reheating device, by eliminating undesired overcooling of the drains in the pipes of a heat transmission pipe bundle. Constitution : A high pressure turbine is driven by main steams from a steam generator. The steams after driving the high pressure turbine are removed with moistures by way of a moisture separator and then re-heated. Extracted steams from the steam generator or the high pressure turbine are used as a heating source for the reheating. In the nuclear turbine power plant having such a constitution, a vessel for separating the drains and the steams resulted from the heat exchange is provided at the outlet of the reheating device and the steams in the vessel are introduced to the inlet of the moisture separator. (Aizawa, K.)

  7. Nuclear power plant

    International Nuclear Information System (INIS)

    Purpose: To suppress corrosion at the inner surfaces of equipments and pipeways in nuclear power plants. Constitution: An injection device comprising a chemical injection tank and a plunger type chemical injection pump for injecting hydrazine as an oxygen remover and ammonia as a pH controller is disposed to the downstream of a condensate desalter column for primary coolant circuits. Since dessolved oxygen in circuit water injected with these chemicals is substantially reduced to zero and pH is adjuted to about 10 - 11, occurrence of stress corrosion cracks in carbon steels and stainless steels as main constituent materials for the nuclear power plant and corrosion products are inhibited in high temperature water, and of corrosion products are inhibited from being introduced as they are through leakage to the reactor core, by which the operators' exposure does can be decreased significantly. (Sekiya, K.)

  8. Physics of BWR MOX fuel results of an international benchmark study by the OECD/NEA nuclear science committee

    International Nuclear Information System (INIS)

    The results of a theoretical benchmark of boiling water reactor (BWR) assembly containing MOX fuel rods are summarised. This study was carried out by the OECD/NEA Working Party on Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR). A modern 10 x 10 BWR design with large internal water structure was chosen for this exercise. It corresponds to an ATRIUM 10 (10-9Q) type with symmetrical water gaps. About 30 solutions were submitted by approximately 20 participants using a dozen different code systems with data from well-known state-of-the-art evaluated nuclear data files, a response which underlines the widespread interest in BWR MOX physics. The discrepancies between the participants for the infinite multiplication factor from beginning of life through burn-ups up to 50 MWd/kg are relatively small (less than 1%). The effect due to diverse evaluated data libraries, e.g. JEF and ENDF represents about 1%. The peaking factor is a local value, more dependent on the methods used in the codes, and with lower compensation effects than for reactivity. The discrepancies are larger in value and there are inconsistencies in the location of the peak. The average values with and without the extreme values differ by 2%, implying that the extreme values could be outside the acceptable range. Other parameters examined include the behaviour of the peaking factor under cold conditions, the evolution of peaking factor with burn-up and the effect of voiding the assembly. Close attention was also paid to the depletion behaviour of gadolinia and the burn-up evolution of the heavy metals. The paper describes the results from this benchmark study and draws conclusions on the consistency of the different solutions provided and provides recommendations for the most effective methods. (author)

  9. Development of evaluation tool for radiation dose rate distribution in PCV of Hamaoka BWR plants based on water chemistry

    International Nuclear Information System (INIS)

    We have developed an evaluation tool for the radiation dose rate distribution of the work areas in the primary containment vessel (PCV) of Units 3, 4 (BWR5) and 5 (ABWR) at Hamaoka NPS. This tool has been constructed based on the transport behavior of radioactive corrosion products in the primary cooling water of BWR. This tool can be used to evaluate quantitatively the effects of the dose reduction methods by water chemistry control or radiation management. It is composed of two calculation codes; water chemistry code (ACTTUBE) and radiation dose rate code (RADTUBE). ACTTUBE calculates the piping dose rates based on the mass balance of corrosion products, 6 kinds of metal and 5 kinds of radionuclide, among the parts of primary cooling water, such as reactor water, feed water, fuel rod surface and out-of-core piping surface. RADTUBE calculates the dose rate distribution based on the radiation shielding calculation from a calculation result of ACTTUBE. Additionally, this tool has a visualization function of calculated radiation dose rate distribution in the PCV by using a wireless controller and 3D glasses/monitor in order to improve user convenience. The accuracy of the tool's calculation results was evaluated using the water chemistry data and radiation dose rate data of the Hamaoka plants. As a result, it was confirmed that this tool had sufficient accuracy to be used in the evaluation of radiation dose rates for the radiation management of actual plants. (author)

  10. Present situation of floating nuclear power plants

    International Nuclear Information System (INIS)

    The present situation of investigation and the future problems of floating nuclear power plants in Japan are examined, referring to those in USA. The committee report on a new power generation system in 1970 is quoted. In this report, the site conditions are supposed to be 5 km offshore, 100m water depth, 60 m/sec wind velocity, 10 m wave height, 200 m wave length, 12 seconds wave period 0.2 g earthquake acceleration, and 2.5 knots tide current. The semisubmersible hull of double construction 15 m under water is employed. A pair of 1,000,000 kW BWR reactors are utilized. A sea water desalting unit using bleed steam from turbines is installed. The solid radioactive wastes packed in drums are disposed in sea. The design and cost estimation were made. The names of the organizations who have made investigation in this field, namely the Civil Engineering Society, the Sience and Technology Agency and other several centers, are reported. The Chubu Electric Power Company is forwarding its project. Referring to the investigations in USA, the project of Atlantic nuclear power station unit is described. A report of plant design has been submitted by O.P.S. to United States Atomic Energy Commission in 1973. The Coastal Area Facilities Act was instituted in New Jersey in 1973. Although the Atlantic nuclear power station has been postponed, it is the most feasible project. For the realization of a floating nuclear power plant in Japan, investigation must be started on the ground construction that can endure the construction of breakwater in water depth of 14 to 30 meter. (Iwakiri, K.)

  11. Cavitation preventing device in a nuclear power plant

    International Nuclear Information System (INIS)

    Purpose: To prevent the generation of cavitation upon loss of feedwater flow rate in BWR nuclear power plant by reliably and rapidly tripping a recycling pump. Constitution: Two phase streams from a nuclear reactor are divided into main steams and saturated water in a steam drum. The deviation between the corresponding flow rate of the main steams and the feedwater flow rate of the feedwater pump sending condensates to the steam drum, as well as the continuing period of the deviation are monitored. Then, if it is detected that both of the deviation and the continuing period thereof exceed specified levels, the recycling pump feeding the saturated water to the reactor is tripped. In this way, the recycling pump can be tripped rapidly and reliably upon loss of feedwater flow rate, whereby the generation of the cavitation can be prevented and the normal operation of the nuclear power plant can be insured. (Moriyama, K.)

  12. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  13. Monitoring method for nuclear power plant

    International Nuclear Information System (INIS)

    In a BWR type nuclear power plant, a critical value to be allowed in view of the operation is determined relative to a minimum critical power ratio (MCPR) and a linear power density in order to keep fuel integrity. The present invention concerns monitoring for the MCPR. That is, CPR is calculated periodically based on each of process data inputted from the plant. An operation limit value is determined based on calculated CPR and each of process data used for the calculation. The operation limit value is corrected, as a function of the control rod position and the reactor flow rate upon CPR calculation and at present. Corrected operation limit value and a low power range monitor (LPRM) signal inputted from the plant are compared to monitor the minimum CPR, that is MCPR. With such procedures, MCPR can be monitored continuously. Further, since the operation limit value is corrected at an insertion position before and after the control rod withdrawal, the non-linear change of the MCPR due to withdrawal of the control rod can be evaluated appropriately. As a result, there is no more necessary to provide unnecessary margin. (I.S.)

  14. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    This report has collected and compiled the data by December in 1981 concerning performances, equipments and installations of the nuclear power plants in Japan. The data have been modified according to the changes produced after previous publication of 1979 edition including BWR and PWR (JAERI-M 8947) and 1980 edition including PWR (JAERI-M 9629), and extended to cover the new plants developed thereafter. All data have been processed and tabulated with a data processing computer program FREP. Besides this report, user also can refer to 'Data List of Nuclear Power Plant in Japan' through terminals equipped at various places in JAERI using TSS (Time Shearing System) network of FACOM M-200, and the explanation of the usage is given in the Appendix. (author)

  15. BWR-plant simulator and its neural network companion with programming under mat lab environment

    International Nuclear Information System (INIS)

    Stand alone nuclear power plant simulators, as well as building blocks based nuclear power simulator are available from different companies throughout the world. In this work, a review of such simulators has been explored for both types. Also a survey of the possible authoring tools for such simulators development has been performed. It is decided, in this research, to develop prototype simulator based on components building blocks. Further more, the authoring tool (Mat lab software) has been selected for programming. It has all the basic tools required for the simulator development similar to that developed by specialized companies for simulator like MMS, APROS and others. Components simulations, as well as integrated components for power plant simulation have been demonstrated. Preliminary neural network reactor model as part of a prepared neural network modules library has been used to demonstrate module order shuffling during simulation. The developed components library can be refined and extended for further development. (author)

  16. Reliability innovations for AREVA NP BWR fuel

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to light water reactors worldwide, representing today more than 185,000 fuel assemblies on the world market including more than 63,000 fuel assemblies for boiling water reactors (BWRs). ATRIUM trademark 10 fuel assemblies have been supplied to a total of 32 BWR plants worldwide resulting in an operating experience over 20,250 fuel assemblies. ATRIUM trademark 10XP and ATRIUM trademark 10XM are AREVA NP's most recent fuel assembly designs featuring improved fuel utilization and achieving high margins to operating limits while maintaining very good reliability. Nevertheless, fuel failures are still encountered in all modern and advanced fuel assembly designs leading to significant operating limitations or unplanned shutdowns of nuclear power plants. The majority of fuel failures in BWR plants are caused by debris fretting, with PCI induced failures being a second leading cause. AREVA NP runs programs to study these root causes and to develop product solutions as part of the continuous improvement process within the Zero Tolerance for Failure (ZTF) initiative. The focus of the ZTF initiative is to further upgrade BWR fuel assembly reliability to achieve the goal of failure free fuel. In the following, two major product improvements are described that will significantly contribute to this goal: - Improved FUELGUARD trademark Lower Tie Plate - Chamfered Fuel Pellet Design (orig.)

  17. BWR internal cracking issues

    International Nuclear Information System (INIS)

    The regulatory issues associated with cracking of boiling water reactor (BWR) internals is being addressed by the Nuclear Regulatory Commission (NRC) staff and is the subject of a voluntary industry initiative. The lessons learned from this effort will be applied to pressurized water reactor (PWR) internals cracking issues

  18. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant

    International Nuclear Information System (INIS)

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  19. Analysis of Heat Balance on Innovative-Simplified Nuclear Power Plant Using Multi-Stage Steam Injectors

    Science.gov (United States)

    Goto, Shoji; Ohmori, Shuichi; Mori, Michitsugu

    The total space and weight of the feedwater heaters in a nuclear power plant (NPP) can be reduced by replacing low-pressure feedwater heaters with high-efficiency steam injectors (SIs). The SI works as a direct heat exchanger between feedwater from condensers and steam extracted from turbines. It can attain pressures higher than the supplied steam pressure. The maintenance cost is lower than that of the current feedwater heater because of its simplified system without movable parts. In this paper, we explain the observed mechanisms of the SI experimentally and the analysis of the computational fluid dynamics (CFD). We then describe mainly the analysis of the heat balance and plant efficiency of the innovative-simplified NPP, which adapted to the boiling water reactor (BWR) with the high-efficiency SI. The plant efficiencies of this innovative-simplified BWR with SI are compared with those of a 1100MWe-class BWR. The SI model is adopted in the heat balance simulator as a simplified model. The results show that the plant efficiencies of the innovate-simplified BWR with SI are almost equal to those of the original BWR. They show that the plant efficiency would be slightly higher if the low-pressure steam, which is extracted from the low-pressure turbine, is used because the first-stage of the SI uses very low pressure.

  20. Survey of BWR plant personnel on shutdown safety practices and risk management needs. EPRI Outage Risk Assessment and Management (ORAM) program. Final report, March 1992

    International Nuclear Information System (INIS)

    In response to increasing attention to potential risks during shutdown conditions, nuclear utilities are explicitly addressing safety in their outage planning and management. As part of an integrated package to assist utilities, a survey of BWR utilities was conducted. Detailed interviews with three BWR utilities were supplemented by insights from NRC and BWR Owners Group surveys. The primary objective was to acquire information for use in developing shutdown risk management guidance. Summaries of the interviews and other surveys resulted in thirteen detailed conclusions related to 1) current levels of safety reviews that have a bearing on shutdown operations, 2) the current process for outage planning and management, 3) outage management policies and guidelines that are in use, 4) perceived problems and challenges, and 5) recommendations for future risk management guidance. (author)

  1. Nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    The present invention concerns an improvement for corrosion resistance of the welded portion of materials which constitutes a reprocessing plant of spent nuclear fuels. That is, Mo-added austenite stainless steel is used for a plant member at the portion in contact with a nitric acid solution. Then, laser beams are irradiated to the welded portion of the plant member and the surface layer is heated to higher than 1,000degC. If such a heat treatment is applied, the degradation of corrosion resistance of the welded portion can be eliminated at the surface. Further, since laser beams are utilized, heating can be limited only to the surface. Accordingly, undesired thermal deformation of the plant members can be prevented. As a result, the plant member having high pit corrosion resistance against a dissolution solution for spent fuels containing sludges comprising insoluble residue and having resistance to nitric acid solution also in the welded portion substantially equal to that of the matrix can be attained. (I.S.)

  2. Obrigheim nuclear power plant

    International Nuclear Information System (INIS)

    In 1973 the 345 MW pressurized water nuclear power plant at Obrigheim operated on base load, generating approximately 2.63 TWh, approximately 2.5 TWh of which was supplied to the KWO members. The plant availability for the year was 89.9%. Of the 10.1% non-availability, 6.4% (23 d) was caused by refuelling, including inspection, overhaul and repair operations and routine tests carried out in September 1973. 3.3% was due to stoppages for repairs to a steam generator and the two main cooling pumps, while 0.4% resulted from failures in the electrical section of the plant. The plant was shut down seven times in all, including three scrams. The average core burnup at the end of the fourth cycle (1 September 1973) was 18900 MWd/tU, representing an average burnup of approximately 37500 MWd/tU for a fuel element used in all four cycles. The operating performance of the steam generators and the result of the steam generator inspection carried out during refuelling in 1973 suggest no progressive damage. The quantities of radioactive materials released to the environment in 1973 were well below the officially permitted levels. The availability of the plant from the beginning of pilot operation in 1969 to the end of 1973 was 83.7 %

  3. Nuclear Power Plant 1996

    International Nuclear Information System (INIS)

    Again this year, our magazine presents the details of the conference on Spanish nuclear power plant operation held in February and that was devoted to 1996 operating results. The Protocol for Establishment of a New Electrical Sector Regulation that was signed last December will undoubtedly represent a new challenge for the nuclear industry. By clearing stating that current standards of quality and safety should be maintained or even increased if possible, the Protocol will force the Sector to improve its productivity, which is already high as demonstrated by the results of the last few years described during this conference and by recent sectorial economic studies. Generation of a nuclear kWh that can compete with other types of power plants is the new challenge for the Sector's professionals, who do not fear the new liberalization policies and approaching competition. Lower inflation and the resulting lower interest rates, apart from being representative indices of our economy's marked improvement, will be very helpful in facing this challenge. (Author)

  4. Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

    International Nuclear Information System (INIS)

    An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented

  5. Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  6. Operational reliability of risk significant systems in US nuclear power plants

    International Nuclear Information System (INIS)

    This paper summarizes the recent results of a programme being sponsored by the Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data to monitor and assess the performance of certain risk significant systems and components at US commercial nuclear power plants. The results of reliability and engineering analyses of the high pressure coolant injection and high pressure core spray systems for BWR plants, and the high pressure coolant injection and high pressure core spray systems for BWR plants, and the emergency diesel generators for both BWR and PWR plants are presented. The analyses are based on available plant operating experience data that reflects systems actuation and operation in conditions as close to probabilistic risk assessment (PRA) accident conditions as practical. Industry-wide trends and plant specific operational reliability estimates are made using data covering the period 1987-1993. Comparisons of results from this work are made to the reliability estimates derived from several PRAs and individual plant examinations (IPEs). Insights gained from the operating experience assessments are also presented. (author). 8 refs, 3 figs, 6 tabs

  7. Serpent: an alternative for the nuclear fuel cells analysis of a BWR

    International Nuclear Information System (INIS)

    In the last ten years the diverse research groups in nuclear engineering of the Universidad Nacional Autonoma de Mexico and Instituto Politecnico Nacional (UNAM, IPN), as of research (Instituto Nacional de Investigaciones Nucleares, ININ) as well as the personnel of the Nuclear Plant Management of the Comision Federal de Electricidad have been using the codes Helios and /or CASMO-4 in the generation of cross sections (X S) of nuclear fuel cells of the cores corresponding to the Units 1 and 2 of the nuclear power plant of Laguna Verde. Both codes belong to the Studsvik-Scandpower Company who receives the payment for the use and their respective maintenance. In recent years, the code Serpent appears among the nuclear community distributed by the OECD/Nea which does not has cost neither in its use neither in its maintenance. The code is based on the Monte Carlo method and makes use of the processing in parallel. In the Escuela Superior de Fisica y Matematicas of the IPN, the personnel has accumulated certain experience in the use of Serpent under the direction of personal of the ININ; of this experience have been obtained for diverse fuel burned, the infinite multiplication factor for three cells of nuclear fuel, without control bar and with control bar for a known thermodynamic state fixed by: a) the fuel temperature (Tf), b) the moderator temperature (Tm) and c) the vacuums fraction (α). Although was not realized any comparison with the X S that the codes Helios and CASMO-4 generate, the results obtained for the infinite multiplication factor show the prospective tendencies with regard to the fuel burned so much in the case in that is not present the control bar like when it is. The results are encouraging and motivate to the study group to continue with the X S generation of a core in order to build the respective library of nuclear data as a following step and this can be used for the codes PARCS, of USA NRC, DYN3D of HZDR, or others developed locally in the IPN

  8. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  9. Floating nuclear power plants

    International Nuclear Information System (INIS)

    This article examines the legal regime for floating nuclear power plants (FNPs), in view of the absence of specific US legislation and the very limited references to artificial islands in the Law of the Sea Convention. The environmental impacts of FNPs are examined and changes in US regulation following the Three Mile Island accident and recent US court decisions are described. References in the Law of the Sea Convention relevant to FNPs are outlined and the current status of international law on the subject is analysed. (author)

  10. Obrigheim nuclear power plant

    International Nuclear Information System (INIS)

    The gross output of the 345MWe pressurized water nuclear power station at Obrigheim, operation on base load, amounted to about 2.57TWh in 1974, the net power fed to the grid being about 2.44TWh. The core was used to its full capacity until 10 May 1974. Thereafter, the reactor was on stretch-out operation with steadily decreasing load until refuelled in August 1974. Plant availability in 1974 amounted to 92.1%. Of the 7.9% non-availability, 7.87% was attributable to the refuelling operation carried out from 16 August to 14 September and to the inspection, overhaul and repair work and the routine tests performed during this period. The plant was in good condition. Only two brief shutdowns occurred in 1974, the total outage time being 21/2 hours. From the beginning of trial operation in March 1969 to the end of 1974, the plant achieved an availability factor of 85.2%. The mean core burnup at the end of the fifth cycle was 19600 MWd/tonne U, with one fuel element that had been used for four cycles achieving a mean burnup of 39000 MWd/tonne U. The sipping test on the fuel elements revealed defective fuel-rods in a prototype plutonium fuel element, a high-efficiency uranium fuel element and a uranium fuel element. The quantities of radioactive substances released to the environment in 1974 were far below the officially permitted values. In july 1974, a reference preparation made up in the nuclear power station in October 1973 was discovered by outsiders on the Obrigheim municipality rubbish tip. The investigations revealed that this reference preparation had very probably been abstracted from the plant in October 1973 and arrived at the rubbish tip in a most irregular manner shortly before its discovery

  11. Garigliano nuclear power plant

    International Nuclear Information System (INIS)

    During the period under review, the Garigliano power station produced 1,028,77 million kWh with a utilization factor of 73,41% and an availability factor of 85,64%. The disparity between the utilization and availability factors was mainly due to a shutdown of about one and half months owing to lack of staff at the plant. The reasons for nonavailability (14.36%) break down as follows: nuclear reasons 11,49%; conventional reasons 2,81%; other reasons 0,06%. During the period under review, no fuel replacements took place. The plant functioned throughout with a single reactor reticulation pump and resulting maximum available capacity of 150 MWe gross. After the month of August, the plant was operated at levels slightly below the maximum available capacity in order to lengthen the fuel cycle. The total number of outages during the period under review was 11. Since the plant was brought into commercial operation, it has produced 9.226 million kWh

  12. Decommissioning of light-water reactor nuclear power plants

    International Nuclear Information System (INIS)

    This study deals with the technical and economic questions posed by the decommissioning of light-water reactor nuclear power plants of the 900-1300 MW class, account being taken of the distinctions between boiling- and pressurized-water reactors. Possible decommissioning alternatives and the disposal or confinement of activity are discussed. It emerges from the discussion that decommissioning, and even total dismantlement of these nuclear power plants is in principle feasible. The activity inventory, one year after shutdown, is calculated to be about 3 X 107 Ci for the BWR and 4 X 106 Ci for the PWR; 40 years after shutdown these figures are reduced to 2 X 106 and 4 X 105 Ci, respectively. The decommissioning costs to be expected are also estimated. This estimate serves as the basis for an economic comparison by the present worth method. The economic comparison shows that total dismantlement after a cooling time of one year is more than four times as expensive as interim confinement followed by total dismantlement waiting period of 40 years. The present worths for immediate total dismantlement are estimated at DM 200 million for the BWR and DM 170 million for the PWR; for the other alternative, they are put at DM 45 million for the BWR and DM 42 million for the PWR. A still open question is posed by the final storage of the large quantities of bulky radioactive waste arising in partial or total dismantlement. Since no decision on the storage method has yet been taken, disposal in casks is stipulated as a boundary condition in the estimation of the costs, although this is an unrealistic assumption. It is to be presumed that the costs of disposal can be reduced given appropriate final storage. (Auth.)

  13. Physical protection of nuclear facilities. Quarterly progress report, July--September 1978. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, L.D. (ed.)

    1979-01-01

    Major activities during the fourth quarter of FY78 included (1) the vital area analysis of operational reactors and characterization of the Standardized Nuclear Unit Power Plant System (SNUPPS), (2) the algorithm development of a new pathfinding computer code, (3) the completion of contractor-supported work for the component generic data base, (4) the refinement of tests related to human parameters modeling, and (5) the addition of improvements to and demonstration of the Safeguards Automated Facility Evaluation (SAFE), Safeguards Network Analysis Procedure (SNAP), and Fixed-Site Neutralization Model (FSNM) methodologies.

  14. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    International Nuclear Information System (INIS)

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

  15. In-core sipping at BWR plants in only 16 hours

    International Nuclear Information System (INIS)

    Introduction of a new FRAMATOME ANP's fuel sipping system which was first manufactured for Germany's twin-unit plant Gundremmingen. It permits all fuel assemblies in the core to be sipped during a refueling outage. It is a direct follow-on from the sipping hood for eight fuel assemblies already supplied by to various nuclear plants. Depending on the size of the core, the new sipping system is designed for 16 fuel assemblies to be tested simultaneously. The new technique that uses this hood for 16 fuel assemblies, in which the gases obtained from water samples are scanned on-line for 85Kr and 133Xe provides immediate information on fuel integrity. (orig.)

  16. Evaluating and improving nuclear power plant operating performance

    International Nuclear Information System (INIS)

    This report aims to provide the basis for improvements in the understanding of nuclear power plants operation and ideas for improving future productivity. The purpose of the project was to identify good practices of operating performance at a few of the world's most productive plants. This report was prepared through a series of consultants meetings, a specialists meeting and an Advisory Group meeting with participation of experts from 23 Member States. The report is based on self-assessment of half a dozen plants that have been chosen as representatives of different reactor types in as many different countries, and the views and assessment of the participants on good practices influencing plant performance. Three main areas that influence nuclear power plant availability and reliability were identified in the discussions: (1) management practices, (2) personnel characteristics, and (3) working practices. These areas cover causes influencing plant performance under plant management control. In each area the report describes factors or good practices that positively influence plant availability. The case studies, presented in annexes, contain the plant self-assessment of areas that influence their availability and reliability. Six plants are represented in the case studies: (1) Dukovany (WWER, 1760 MW) in the Czech Republic; (2) Blayais (PWR, 3640 MW) in France; (3) Paks (WWER, 1840 MW) in Hungary; (4) Wolsong 1 (PHWR, 600 MW) in the Republic of Korea; (5) Trillo 1 (PWR, 1066 MW) in Spain; and (6) Limerick (BWR, 2220 MW) in the United States of America

  17. Characterization of solidified radioactive wastes produced at Montalto di Castro BWR plant with reference to the site storage

    International Nuclear Information System (INIS)

    The cement solidification of the Montalto di Castro BWR plant radwastes has been studied both from the point of view of the mixtures of formulation and of the product characterization. Five radwaste types and mixtures of them have been taken into consideration, determining the best chemical formulations starting from the compressive strenght as leading parameter. The solidified products have been characterized from the point of view of the freeze and thawing resistance, the water immersion resistance, the leachability, the dimensional changes and the free standing water. All the tests have been performed taking into account the real site conditions, so the leaching tests and the water immersion tests have been carried out using sea water and table water as leachant

  18. Nuclear coupled flow instability study for natural circulation BWR startup transient

    International Nuclear Information System (INIS)

    Natural circulation Boiling Water Reactor (BWR) startup transient was investigated in Purdue University Multidimensional Test Assembly (PUMA) facility based on a natural circulation BWR design. Strategy and results of the experiments, which consider the effects of void-reactivity and fuel heat conduction time constant, are discussed. Total reactivity is treated to be composed of two components: external reactivity due to control rod motion and void-reactivity. A detailed analysis for heat conduction problem is performed to derive dimensionless groups. Based on area-averaged heat conduction equations for pellet and clad regions, Fourier and Biot numbers are derived to simulate wall heat flux response. Power transient, which has been used for startup transient investigation without void-reactivity feedback is used to derive the control rod reactivity. Twelve conductivity probes are used to measure local void fraction inside core at three axial locations. The local void-fraction data is used to calculate volume average void fraction, which is used to calculate the voil-reactivity. A real-time Point Kinetic Model solver is implemented to PUMA heater power control program to determine power transient during startup. The results demonstrate that the inclusion of void-reactivity feedback worsen the scenario for startup instabilities and may cause large amplitude neutron flux oscillations. (author)

  19. Nuclear power plant risk management

    International Nuclear Information System (INIS)

    This book collects papers on nuclear plant risk management. Topics include: basic concepts in risk analysis and decision theory; establishing integrated plant models; core and containment response; and site modeling and consequences

  20. Recent trends of new technologies for BWR power plant instrumentation and control

    International Nuclear Information System (INIS)

    For the improvement of the operation reliability of nuclear power stations and its rational realization, Hitachi Ltd. has advanced research and development with the basic thinking of the harmony of men and machines. This time, the new type synthetic, digital, central monitoring and control system for No. 7 plant in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. was completed, and is being installed at the site. For the purpose of further heightening the operation reliability and maintainability of the plant, advanced man-machine function was positively adopted, and the range of automation was expanded, including the automation of control rod operation, therefore, it is the synthetic digitized system of plantwide. It is considered to show one direction toward the more human friendly measurement and control system of next generation. In this report, as the features of the new type synthetic, digital, central monitoring and control system (NUCAMM-90), the advancement of human interface, the automation of control rod operation and the expansion of the range of automation aiming at the operation system with monitoring as the main, and the digitizing of safety and protection system are described. Efforts are exerted for the research and development of the man-machine system of media fusion type, the support systems for operation planning and the time of abnormality and so on. (K.I.)

  1. Aging management guideline for commercial nuclear power plants-pumps

    International Nuclear Information System (INIS)

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  2. Aging Management Guideline for commercial nuclear power plants: Electrical switchgear

    International Nuclear Information System (INIS)

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of this AMG to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner which allows personnel responsible for performance analysis and maintenance, to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  3. Aging management guideline for commercial nuclear power plants-pumps

    Energy Technology Data Exchange (ETDEWEB)

    Booker, S.; Katz, D.; Daavettila, N.; Lehnert, D. [MDC-Ogden Environmental and Energy Services, Southfield, MI (United States)

    1994-03-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  4. A study on maintenance optimization by the automatic planning tool for regular plant outage work in nuclear power plant using the logic programming language 'Prolog'

    International Nuclear Information System (INIS)

    This paper discusses maintenance optimization by the automatic planning tool for regular plant outage work in nuclear power plant using the logic programming language 'Prolog'. As a result of consideration, the following results were obtained. (1) The automatic planning tool for regular plant outage in nuclear power plant was developed. (2) Using this tool, the work plan for BWR primary recirculation system and residual heat removal system was automatically made on the condition of flattening man loading over the plant outage schedule as much as possible. (3) Several points for improving the developed tool were listed. (author)

  5. An overview of the BWR ECCS strainer blockage issues

    Energy Technology Data Exchange (ETDEWEB)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-03-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, {open_quotes}Containment Emergency Sump Performance,{close_quotes} and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts.

  6. Nuclear power plant operator licensing

    International Nuclear Information System (INIS)

    The guide applies to the nuclear power plant operator licensing procedure referred to the section 128 of the Finnish Nuclear Energy Degree. The licensing procedure applies to shift supervisors and those operators of the shift teams of nuclear power plant units who manipulate the controls of nuclear power plants systems in the main control room. The qualification requirements presented in the guide also apply to nuclear safety engineers who work in the main control room and provide support to the shift supervisors, operation engineers who are the immediate superiors of shift supervisors, heads of the operational planning units and simulator instructors. The operator licensing procedure for other nuclear facilities are decided case by case. The requirements for the basic education, work experience and the initial, refresher and complementary training of nuclear power plant operating personnel are presented in the YVL guide 1.7. (2 refs.)

  7. Nuclear power plant

    International Nuclear Information System (INIS)

    Purpose: To prevent liquid wastes from being discharged out of the system by processing to recover them in the nuclear reactor and reusing them. Constitution: Discharge of liquid wastes to the surrounding atmosphere are completely eliminated by collecting floor drains, a part of processing water for the regeneration of liquid wastes, non-radioactive steam drains and laundry drains conventionally discharged so far out of the system, processing them in a concentrator, a desalter or the like into water of a high purity and extremely low radioactive concentration, storing the water in an exclusive storage tank and supplying it as a steam or supplementing water to each portion in the plant that requires water of such high purity and extremely low radioactivity. (Yoshihara, H.)

  8. Study and characterization of noble metal deposits on similar rusty surfaces to those of the reactor U-1 type BWR of nuclear power station of Laguna Verde

    International Nuclear Information System (INIS)

    In the present investigation work, were determined the parameters to simulate the conditions of internal oxidation reactor circulation pipes of the nuclear power plant of Laguna Verde in Veracruz. We used 304l stainless steel cylinders with two faces prepared with abrasive paper of No. 600, with the finality to obtain similar surface to the internal circulation piping nuclear reactor. Oxides was formed within an autoclave (Autoclave MEX-02 unit B), which is a device that simulates the working conditions of the nuclear reactor, but without radiation generated by the fission reaction within the reactor. The oxidation conditions were a temperature of 280 C and pressure of 8 MPa, similar conditions to the reactor operating in nuclear power plant of Laguna Verde in Veracruz, Mexico (BWR conditions), with an average conductivity of 4.58 ms / cm and 2352 ppb oxygen to simulate normal water chemistry NWC. Were obtained deposits of noble metal oxides formed on 304l stainless steel samples, in a 250 ml autoclave at a temperature range of 180 to 200 C. The elements that were used to deposit platinum-rhodium (Pt-Rh) with aqueous Na2Pt (OH)6 and Na3Rh (NO2)6, Silver (Ag) with an aqueous solution of AgNO3, zirconium (Zr) with aqueous Zr O (NO3) and ZrO2, and zinc (Zn) in aqueous solution of Zn (NO3)2 under conditions of normal water chemistry. Also there was the oxidation of 304l stainless steel specimens in normal water chemistry with a solution of Zinc (Zn) (NWC + Zn). Oxidation of the specimens in water chemistry with a solution of zinc (Zn + NWC) was prepared in two ways: within the MEX-02 autoclave unit A in a solution of zinc and a flask at constant temperature in zinc solution. The oxides formed and deposits were characterized by scanning electron microscopy, energy dispersive X-ray analysis, elemental field analysis and X-ray diffraction. By other hand was evaluated the electrochemical behavior of the oxides formed on the surface of 304l stainless steel in normal water

  9. Nuclear power plant

    International Nuclear Information System (INIS)

    The present invention provides a nuclear power plant which generates thermoelectric power by utilizing heat generated by fission reaction. Namely, a fuel/thermoelectric material is made of a semiconductor material containing fission products or a semimetal material containing fission products. A reactor container contains the fuel/thermoelectric material and a reactor core constituted by the fuel/thermoelectric material. The reactor container comprises coolants for removing heat generated by nuclear reaction of fission products from the reactor core and a high temperature side electrode connected to a central portion of the fuel/thermoelectric material and a low temperature side electrode connected to the outside of the fuel/thermoelectric material. Electromotive force is caused in the fuel/thermoelectric material by temperature difference upon combustion caused at the central portion and the outer surface of the fuel/thermoelectric material. The electromotive force is taken out of the high temperature side electrode and the low temperature side electrode. (I.S.)

  10. Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Analia Bonelli

    2012-01-01

    Full Text Available A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.

  11. Operate a Nuclear Power Plant.

    Science.gov (United States)

    Frimpter, Bonnie J.; And Others

    1983-01-01

    Describes classroom use of a computer program originally published in Creative Computing magazine. "The Nuclear Power Plant" (runs on Apple II with 48K memory) simulates the operating of a nuclear generating station, requiring students to make decisions as they assume the task of managing the plant. (JN)

  12. Management of radioactive waste from a major core damage in a BWR power plant

    International Nuclear Information System (INIS)

    Large amounts of fission products would be released in case of a major core damage in a nuclear power reactor. In this theoretical study the core damage is caused by a loss of coolant accident followed by a complete loss of all electric power for about 30 minutes resulting in the release of 10% of the core inventory of noble gases. A second case has also been briefly studied, in which the corresponding core damage is supposed to be created merely by the complete loss of electric power during a limited time period. It appears from the study that the radioactive waste generated as a consequence of an accident of the extent can be managed in the reference reactor with only minor modifications required in the waste plant. The detailed results of the study are reactor specific, but many of the findings and recommendations are generally applicable. (author) 28 refs

  13. Owners of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hudson, C.R.; White, V.S.

    1996-11-01

    Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of July 1996. The report is divided into sections representing different aspects of nuclear plant ownership.

  14. Owners of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Reid, R.L.

    2000-01-12

    Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of November 1999. The report is divided into sections representing different aspects of nuclear plant ownership.

  15. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  16. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    The list indicates percentage ownership of commercial nuclear power plants by utility companies as of September 1, 1982. The list includes all plants licensed to operate, under construction, docketed for NRC safety and environmental reviews, or under NRC antitrust review. Part I lists plants alphabetically with their associated applicants and percentage ownership. Part II lists applicants alphabetically with their associated plants and percentage ownership. Part I also indicates which plants have received operating licenses

  17. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    International Nuclear Information System (INIS)

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner's Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section

  18. Experience for plant monitoring design in Italian BWR NPP and future trends in man-machine interface

    International Nuclear Information System (INIS)

    TMI accidental sequence and daily-gained operating experience on italian and abroad NPPs have affected in depth the approach to the design of information presentation to the Control Room staff. It has been cleared that most problems in plant operation arise from a poor and inadequate information system. The main lacks have been identified in the Control Room lay-out and information organization. This has pushed designers both to improve the Control Room environment and to better exploit the computer data processing and data presentation capabilities. The paper deals with the basic criteria for the design and the design review of a computerized system to be inserted in a hybrid Control Room in Italian 981 Mwe BWR-6 NPP, where the concepts outlined above were taken-up from the very beginning. The Control Room keeps conventional instrumentation arranged in a human-factor lay-out, according to post-TMI requirements, and adds a powerful computer-based information system for advanced alarm presentation and plant supervision during both normal and emergency conditions with high data reliability. Colour videounits and operating panels are functionally integrated to create powerful operator work-stations. Emphasis is mostly given on the revision work for video-unit displays and Man-System Communication carried out in cooperation with Halden Reactor Project human factor and plant operation experts. The work peculiarity has been a strong care on the integration between conventional and computerized information presentation, with particular regard to common information and code consistency. (author)

  19. World wide nuclear plant performance

    International Nuclear Information System (INIS)

    The objectives of the paper are to compare the performance of different reactor systems, to identify the determinants of plant performance, to examine the evidence of technological maturation, to discover the principal causes of outage or unavailability and to draw implications for technology policy. Judgments are neither made nor implied about the relative merits of nuclear and fossil fuel plants, nor on safety issues. This study covers all commercial nuclear plants installed in the non-communist countries. (author)

  20. Nuclear Security for Floating Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Skiba, James M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scherer, Carolynn P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-13

    Recently there has been a lot of interest in small modular reactors. A specific type of these small modular reactors (SMR,) are marine based power plants called floating nuclear power plants (FNPP). These FNPPs are typically built by countries with extensive knowledge of nuclear energy, such as Russia, France, China and the US. These FNPPs are built in one country and then sent to countries in need of power and/or seawater desalination. Fifteen countries have expressed interest in acquiring such power stations. Some designs for such power stations are briefly summarized. Several different avenues for cooperation in FNPP technology are proposed, including IAEA nuclear security (i.e. safeguards), multilateral or bilateral agreements, and working with Russian design that incorporates nuclear safeguards for IAEA inspections in non-nuclear weapons states

  1. Nuclear Security for Floating Nuclear Power Plants

    International Nuclear Information System (INIS)

    Recently there has been a lot of interest in small modular reactors. A specific type of these small modular reactors (SMR,) are marine based power plants called floating nuclear power plants (FNPP). These FNPPs are typically built by countries with extensive knowledge of nuclear energy, such as Russia, France, China and the US. These FNPPs are built in one country and then sent to countries in need of power and/or seawater desalination. Fifteen countries have expressed interest in acquiring such power stations. Some designs for such power stations are briefly summarized. Several different avenues for cooperation in FNPP technology are proposed, including IAEA nuclear security (i.e. safeguards), multilateral or bilateral agreements, and working with Russian design that incorporates nuclear safeguards for IAEA inspections in non-nuclear weapons states

  2. Pre service and inservice inspection of light water moderated nuclear power plant pressure-containing components

    International Nuclear Information System (INIS)

    The authors comment the current national (United States) and international techniques, experience and problems in the field of pre-service and in-service inspection of pressure vessels and piping of nuclear power plant reactors. They also comment the personnel qualification issue, and results obtained during inspections of components of various nuclear plant reactor vessel (BWR and PWR). They highlight and comment the major problems associated with in-service inspection: the scarcity of personnel trained and qualified for reactor system examination, the variability of procedures and data recording, and the problem of inspection personnel radiation exposure

  3. Elecnuc. Nuclear power plants worldwide

    International Nuclear Information System (INIS)

    This small folder presents a digest of some useful information concerning the nuclear power plants worldwide and the situation of nuclear industry at the end of 1997: power production of nuclear origin, distribution of reactor types, number of installed units, evolution and prediction of reactor orders, connections to the grid and decommissioning, worldwide development of nuclear power, evolution of power production of nuclear origin, the installed power per reactor type, market shares and exports of the main nuclear engineering companies, power plants constructions and orders situation, evolution of reactors performances during the last 10 years, know-how and development of nuclear safety, the remarkable facts of 1997, the future of nuclear power and the energy policy trends. (J.S.)

  4. An engineer-constructor's view of nuclear power plant safety

    International Nuclear Information System (INIS)

    At SWEC we have been involved in the development of safety features of nuclear power plants ever since we served as the engineer-constructur for the first commerical nuclear power station at Shippingport, Pennsylvania, in the 1950s. Our personnel have pioneered a number of safety innovations and improvements. Among these innovations is the subatmospheric containment for pressurized water reactor (PWR) power plants. This type of containment is designed so that leakage will terminate within 1 to 2 hours of the worst postulated loss of coolant accident. Other notable contributions include first use of reinforced-concrete atmospheric containments for PWR power plants and of reinforced-concrete, vapor-suppression containments for boiling water reactor (BWR) power plants. Both concepts meet rigorous U.S. safety requirements. SWEC has performed a substantial amount of work on developing standardized plant designs and has developed standardized engineering and construction techniques and procedures. Standardization concepts are being developed in Canada, France, USSR, and Germany, as well as in the United States. The West German convoy concept, which involves developing a number of standardized plants in a common effort, has been quite successful. We believe standardization contributes to safety in a number of ways. Use of standardized designs, procedures, techniques, equipment, and methods increases efficiency and results in higher quality. Standardization also reduces the design variations with which plant operators, emergency teams, and regulatory personnel must be familiar, thus increasing operator capability, and permits specialized talents to be focused on important safety considerations. (orig./RW)

  5. EMP and nuclear plant safety

    International Nuclear Information System (INIS)

    The electromagnetic pulse (EMP) from a high-altitude nuclear detonation consists of a transient pulse of high-intensity electromagnetic fields that induce current and voltage transients in electrical conductors. Although most nuclear power-plant cables are not directly exposed to these fields, the attenuated EMP fields that propagate into the plant will couple some EMP energy to these cables. The article attempts to predict the probable effects of the EMP transients that could be induced in critical circuits of safety-related systems. It is concluded that the most likely consequence of EMP for nuclear plants is an unscheduled shutdown. In general, EMP could be a nuisance to nuclear power plants, but it is not considered a serious threat to plant safety

  6. Technologies for nuclear plant decommissioning

    International Nuclear Information System (INIS)

    After the commercial operation of a nuclear power plant has been shutdown, the plant enters a decommissioning phase where it is dismantled and removed. The Tokai Power Station was shutdown at the end of March 1998, followed by 'Fugen' and a light water reactor. The number of decommissioned plants in Japan is likely to increase in the future. Based on experience gained from the construction and maintenance of nuclear plants, Fuji Electric has developed techniques essential for decommissioning work. This paper describes recent technologies developed in this field, such as remote dismantling techniques for the reactor core and treatment and disposal techniques for the dismantled waste. (author)

  7. Nuclear Power Plant Simulation Game.

    Science.gov (United States)

    Weiss, Fran

    1979-01-01

    Presents a nuclear power plant simulation game which is designed to involve a class of 30 junior or senior high school students. Scientific, ecological, and social issues covered in the game are also presented. (HM)

  8. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  9. Technology and costs for decommissioning of Swedish nuclear power plants

    International Nuclear Information System (INIS)

    The decommissioning study for the Swedish nuclear power plants has been carried out during 1992 to 1994 and the work has been led by a steering group consisting of people from the nuclear utilities and SKB. The study has been focused on two reference plants, Oskarshamn 3 and Ringhals 2. Oskarshamn 3 is a boiling water reactor (BWR) and Ringhals 2 is a pressurized water reactor (PWR). Subsequently, the result from these plants have been translated to the other Swedish plants. The study gives an account of the procedures, costs, waste quantities and occupational doses associated with decommissioning of the Swedish nuclear power plants. Dismantling is assumed to start immediately after removal of the spent fuel. No attempts at optimization, in terms of technology or costs, have been made. The nuclear power plant site is restored after decommissioning so that it can be released for use without restriction for other industrial activities. The study shows that a reactor can be dismantled in about five years, with an average labour force of about 150 persons. The maximum labour force required for Oskarshamn 3 has been estimated to about 300 persons. This peak load occurred the first years but is reduced to about 50 persons during the demolishing of the buildings. The cost of decommissioning Oskarshamn 3 has been estimated to be about MSEK 940 in January 1994 prices. The decommissioning of Ringhals 2 has been estimated to be MSEK 640. The costs for the other Swedish nuclear power plants lie in the range MSEK 590-960. 17 refs, 21 figs, 15 tabs

  10. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR

    International Nuclear Information System (INIS)

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  11. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics. (author)

  12. Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR

    International Nuclear Information System (INIS)

    The stress corrosion cracking susceptibility was studied in sensitized, solution annealed 304 steel, and in 304-L welded with a heat treatment that simulated the radiation induced segregation, by the slow strain rate test technique, in a similar environment of a boiling water reactor (BWR), 288 C, 8 MPa, low conductivity and a electrochemical corrosion potential near 200 mV. vs. standard hydrogen electrode (She). The electrochemical noise technique was used for the detection of the initiation and propagation of the cracking. The steels were characterized by metallographic studies with optical and scanning electronic microscopy and by the electrochemical potentiodynamic reactivation of single loop and double loop. In all the cases, the steels present delta ferrite. The slow strain rate tests showed that the 304 steel in the solution annealed condition is susceptible to transgranular stress corrosion cracking (TGSCC), such as in a normalized condition showed granulated. In the sensitized condition the steel showed intergranular stress corrosion cracking, followed by a transition to TGSCC. The electrochemical noise time series showed that is possible associated different time sequences to different modes of cracking and that is possible detect sequentially cracking events, it is means, one after other, supported by the fractographic studies by scanning electron microscopy. The parameter that can distinguish between the different modes of cracking is the re passivation rate, obtained by the current decay rate -n- in the current transients. This is due that the re passivation rate is a function of the microstructure and the sensitization. Other statistic parameters like the localized index, Kurtosis, Skew, produce results that are related with mixed corrosion. (Author)

  13. Nuclear power plant life extension

    International Nuclear Information System (INIS)

    Current licensing emphasis is different from that existing when older nuclear power plants received their operating licenses. Today, there is increased regulatory focus on utility performance measured through new inspections such as the safety system functional inspection and the safety systems outage modifications inspection; and the use of performance indicators. These programs are intended to identify poorer performing plants and concentrate on improving overall plant performance. This paper discusses how, in an attempt to improve the maintenance performance of the nuclear industry, the U.S. Nuclear Regulatory Commission recently began work on a maintenance rule. In addition, the NRC will require partial probabilistic risk evaluations for each plant that will provide new insights as to a plant's risk. Risk information is being used to prioritize inspections

  14. Investigation of BWR [boiling water reactor] instability phenomena using RETRAN-03

    International Nuclear Information System (INIS)

    In 1988, LaSalle, a boiling water reactor (BWR)/5, experienced severe flux oscillations following a trip of both recirculation pumps. The flux oscillations were terminated by an automatic scram at 118% of rated neutron flux. As a result of this event, the U.S. Nuclear Regulatory Commission has asked the BWR utilities to develop procedural or hardware changes that will assure protection of all safety limits. The rapid growth of the oscillations at LaSalle, and the fact that previous stability analyses had predicted the plant to be very stable, emphasizes that a better understanding of this phenomenon is needed before the success of the long-term fixes can be assured. The intent of the Electric Power Research Institute's work was to use BWR transient methods to model reactor instabilities and investigate the factors that dominate this phenomenon. The one-dimensional transient code RETRAN-03 (Ref. 1) was used. The following conclusions are drawn: (1) RETRAN has demonstrated the ability to model BWR instability (nonlinear oscillations). (2) The general system behavior predicted by RETRAN in BWR stability analyses matches theoretical prediction and plant data. (3) These one-dimensional, time-domain results have increased the understanding of BWR stability phenomena and have helped optimize the long-term solutions being developed by the utilities

  15. Reflection of the operating experiences to the nuclear plant design

    International Nuclear Information System (INIS)

    After the first commercial operation in 1970, operating nuclear light water reactors in Japan increase, and, at the end of September 1991, the number of the plants reaches 39 and the total capacity becomes 31890MWe. According to the increase of the nuclear power plants, public requirement for the reliability and safety is getting intensified. The development of Japanese nuclear power plants is basically started by introducing the engineering and technology from the United States of America and various methods for Quality Assurance (QA) are also imported from the US at the same time. After the cumulation of experiences related to the plant construction and operation, including unexpected troubles, the standards for the QA of the design, construction and operating stages and the actual methods of the QA are established in accordance with the cooperative efforts of the Government, electrical power companies and plant suppliers. As the results of those efforts, the capacity factor and the frequency of unplanned shutdown of the nuclear plants have become top levels in the world. It shows that the those activities taken for the improvement of the QA has been effective. The QA activities seem to get into the new turn in recent years. In this report, we present the content of the Japanese national programs such as the Improvement and Standardization of LWR and the Sophistication of LWR, as an example of the movement of reflecting the various operating experiences to the BWR plant design. A few issues are also presented for visualization of the activities. In addition, we present the actual QA activities for confirming the reflected new plant design based on Hitachi's practices, as a sample

  16. Environmental tests of electric cables for nuclear power plants

    International Nuclear Information System (INIS)

    The evaluation of the materials for flame retardation of cables for nuclear power plants and the tests of their environmental resistivity are reported. Several cables to meet the severe requirements for these cables have been produced for trial, which have been subjected to the simulated environment test including Loss of Coolant Accident. The results are summarized as follows: (a) the ethylene-proylene rubber-insulated, chloroprene-sheathed cable has the sufficient properties to use in BWR containment vessels, such as radiation resistance and flame resistance. (b) It is the same in the chloroprene-sheathed cable, that has glass braids on the outermost layer and around each core, and the insulation layer of the silicone rubber electrically stable but mechanically low in radiation resistivity. (c) The flame-resistant cross-linking polyethylene-insulated, low hydrochloric acid vinyl-sheathed cable is usable near but outside BWR containment vessels. (d) The flame-resistant cross-linking polyethylene-insulated, chloroprene-sheathed cable having the insulating layer of polyethylene for each core can be used in PWR containment vessels. (e) Chloroprene and polyvinyl chloride, the flame-resistant sheath materials, deteriorate owing to heat aging alone. Only the insulating layers mainly contribute to the environmental resistivity of cables. (f) The environmental conditions are severer when heat aging is carried out prior to irradiation. The influence of dose rate is rather little. (Wakatsuki, Y.)

  17. Reducing BWR O and M costs through on-line performance monitoring

    International Nuclear Information System (INIS)

    Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants

  18. Maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    This Safety Guide provides detailed guidance on the provisions of the Code on the Safety in Nuclear Power Plants: Operation, IAEA Safety Series No. 50-C-O(Rev.1) on the maintenance of structures, systems and components. Like the Code, the Guide forms part of the IAEA's programme, referred to as the NUSS programme, for establishing Codes and Safety Guides relating to nuclear power plants. Effective maintenance is essential for safe operation of a nuclear power plant. It not only ensures that the level of reliability and effectiveness of all plant structures, systems and components having a bearing on safety remains in accordance with design assumptions and intent, but also that the safety status of the plant is not adversely affected after commencement of operation. Nuclear power plant maintenance requires special attention because of: Limitations set by requirements that a minimum number of components remain operable even when the plant is shut down in order to ensure that all necessary safety functions are guaranteed; Difficulty of access to some plant items even when the plant is shut down, due to radiation protection constraints; Potential radiological hazards to site personnel and the public. This Guide covers the organizational and procedural aspects of maintenance but does not give detailed technical advice on the maintenance of particular plant items. It gives guidance on preventive and remedial measures necessary to ensure that all structures, systems and components important to safety are capable of performing as intended. The Guide covers the organizational and administrative requirements for establishing and implementing preventive maintenance schedules, repairing defective plant items, selecting and training maintenance personnel, providing maintenance facilities and equipment, procuring stores and spare parts, reviewing, controlling and carrying out plant modifications, and generating, collecting and retaining maintenance records for establishing and

  19. Loviisa nuclear power plant analyzer

    International Nuclear Information System (INIS)

    The APROS Simulation Environment has been developed since 1986 by Imatran Voima Oy (IVO) and the Technical Research Centre of Finland (VTT). It provides tools, solution algorithms and process components for use in different simulation systems for design, analysis and training purposes. One of its main nuclear applications is the Loviisa Nuclear Power Plant Analyzer (LPA). The Loviisa Plant Analyzer includes all the important plant components both in the primary and in the secondary circuits. In addition, all the main control systems, the protection system and the high voltage electrical systems are included. (orig.)

  20. Benchmarking of transient codes against cycle 19 STABILITY measurements at Leibstadt nuclear power plant (KKL) - 131

    International Nuclear Information System (INIS)

    Coupled neutronics-thermal hydraulic codes are used by many utilities, research institutes and regulatory authorities worldwide for performing BWR stability analysis. RAMONA-3 has been established in the industry for quite a long time as a reliable time-domain dynamic code with best performance for predictive calculations. Next generation of codes such as RAMONA-5, SIMULATE-3K and POLCA-T, with advanced two-group neutronics and more detailed plant description and thermal hydraulics models have been introduced. The performance of these codes against the stability measurements performed in cycle 19 at the Swiss nuclear power plant Leibstadt (KKL), a BWR/6 from General Electric, is presented in this paper. Important suppliers of the nuclear industry such as Westinghouse Electric Sweden, AREVA NP Germany, Studsvik Scandpower Inc. USA, and the Swiss research institute PSI have participated in this work. The validation of calculation methods against the KKL stability measurements was considered important by the various organizations for different reasons. Amongst others, Studsvik Scandpower aimed at filling a gap in the SIMULATE-3K stability benchmark database to include a jet pumps driven plant, AREVA NP had to fulfill fuel licensing requirements, and Westinghouse planned to launch POLCA-T parallel to a validation of RAMONA-5 as a production code. PSI cooperated with KKL in stability issues from the very beginning and introduced the stability test project in the framework of NACUSP, a European consortium that aimed for a better understanding of the BWR stability problem. For that purpose, this validation provides an assessment of advanced stability codes for modern BWR core designs. (authors)

  1. TVA's nuclear power plant experience

    International Nuclear Information System (INIS)

    This paper reviews TVA's nuclear power plant design and construction experience in terms of schedule and capital costs. The completed plant in commercial operation at Browns Ferry and six additional plants currently under construction represent the nation's largest single commitment to nuclear power and an ultimate investment of $12 billion by 1986. The presentation is made in three separate phases. Phase one will recapitulate the status of the nuclear power industry in 1966 and set forth the assumptions used for estimating capital costs and projecting project schedules for the first TVA units. Phase two describes what happened to the program in the hectic early 1979's in terms of expansion of scope (particularly for safety features), the dramatic increase in regulatory requirements, vendor problems, stretchout of project schedules, and unprecedented inflation. Phase three addresses the assumptions used today in estimating schedules and plant costs for the next ten-year period

  2. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  3. An A BWR demonstration simulator for training and developing technical staff

    International Nuclear Information System (INIS)

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  4. An A BWR demonstration simulator for training and developing technical staff

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  5. An estimation of population doses from a nuclear power plant during normal operation

    International Nuclear Information System (INIS)

    A model is presented for estimation of the potential submersion and inhalation radiation doses to people located within a distance of 1000 km from a nuclear power plant during normal operation. The model was used to calculate doses for people living 200-1000 km from hypothetical nuclear power facility sited near the geographical centre of Denmark. Two kinds of sources are considered for this situation: - unit release of 15 isotopes of noble gases and iodines, - effluent releases from two types of 1000 MWe Light Water Power Reactors: PWR and BWR. Parameter variations were made and analyzed in order to obtain a better understanding of the mechanisms of the model. (author)

  6. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

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

  7. BWR AXIAL PROFILE

    Energy Technology Data Exchange (ETDEWEB)

    J. Huffer

    2004-09-28

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

  8. Fukushima, two years later, modification requirements in nuclear power plants

    International Nuclear Information System (INIS)

    The occurred events in the nuclear power plant of Fukushima Daiichi as consequence of the strong earthquake of 9 grades in the Richter scale and the later tsunami with waves estimated in more than 14 meters high began a series of important questions about the safety of the nuclear power plants in operation and of the new designs. Firstly, have allowed to be questioned on the magnitudes and consequences of the extreme external natural events; that can put in risk the integrity of the safety barriers of a nuclear power plant when being presented in a multiple way. As consequence of the events of the Fukushima Daiichi NPP, the countries with NPPs in operation and /or construction carried out evaluations about their safety operation. They have also realized evaluations about accidents and their impact in the safety, analysis and studies too that have forced to the regulatory bodies to continue a systematic and methodical revision of their procedures and regulations, to identify the possible improvements to the safety in response to the events happened in Japan; everything has taken it to determine the necessity to incorporate additional requirements to the nuclear power plants to mitigate events Beyond the Design Base. Due to Mexico has the nuclear power plant of Laguna Verde, with two units of BWR-5 type with contention Mark III, some the modifications can be applicable to these units to administrate and/or to mitigate the consequences of the possible occurrence of an accident Beyond the Design Base and that could generate a severe accident. In this work an exposition is presented on the modification requirements to confront external natural events Beyond the Design Base, and its application in our country. (Author)

  9. Nuclear power plant performance objectives for the 1990s

    International Nuclear Information System (INIS)

    The development of nuclear energy in recent decades as an indispensable energy source has produced a very high standard of reliability and availability. Light water reactors (PWR and BWR) constitute the basis for the commercial application of nuclear energy. Under construction at the end of 1986 were 133 units with a total capacity of c. 118,000 MW(e). Most of these units will start commercial operation between the end of the 'eighties and middle of the 'nineties. A low level increase rate of approx. 28,000 - 30,000 MW up to the year 2000 may be assumed additionally. The objectives for the 1990s are the backfitting of current LWRs as far as necessary and feasible and as a future task the development of the next generation of LWRs. The main goal is seen in cost reduction and shortening construction time to increase the economic viability of nuclear energy compared with fossil fired power plants. In the technical field the reactor core, nuclear fuel, instrumentation and control are areas for improvement. The nuclear fuel cycle will be optimized for higher burnup, lower uranium utilization by application of modified core management methods, improvements in operating flexibility and saving of natural uranium and future development of reprocessing technology and recycling of reprocessed U and Pu. Instrumentation and control systems are areas for evolutionary modifications and/or changes. With the introduction of computer systems in various fields and in the control room, the development of fibre-optical systems and the introduction of microprocessors higher reliability is possible for control of the entire plant and reduction of man-machine interfaces. The application of part task simulators and special simulators beside the well proven full scope simulator enables optimal personnel training and retraining. Another important area for future development and increased plant availability is reduction of planned and unplanned plant outages, shortening refuelling periods

  10. Robots for nuclear power plants

    International Nuclear Information System (INIS)

    In many industrial applications of robots, the objective is to replace human workers with machines that are more productive, efficient, and accurate. But for nuclear applications, the objective is not so much to replace workers as it is to extend their presence - for example, to project their reach into areas of a nuclear plant where the thermal or radiation environment prohibits or limits a human presence. The economic motivation to use robots for nuclear plant inspection and maintenance is centered on their potential for improving plant availability; a by-product is the potential for reducing the occupational radiation exposure of plant personnel. Robotic equipment in nuclear applications may be divided into two broad categories: single-purpose devices with limited ability to perform different operations, and reprogrammable, multi-purpose robots with some degree of computer-based artificial intelligence. Preliminary assessments of the potential for applying robotics in nuclear power plants - mainly at surveillance and inspection tasks - have been carried out. Future developments are considered

  11. World nuclear power plant capacity

    International Nuclear Information System (INIS)

    This report provides the background information for statistics and analysis developed by NUKEM in its monthly Market Report on the Nuclear Fuel Cycle. The assessments in this Special Report are based on the continuous review of individual nuclear power plant projects. This Special Report begins with tables summarizing a variety of nuclear power generating capacity statistics for 1990. It continues with a brief review of the year's major events regarding each country's nuclear power program. The standard NUKEM Market Report tables on nuclear plant capacity are given on pages 24 and 25. Owing to space limitations, the first year shown is 1988. Please refer to previous Special Reports for data covering earlier years. Detailed tables for each country list all existing plants as well as those expected by NUKEM to be in commercial operation by the end of 2005. An Appendix containing a list of abbreviations can be found starting on page 56. Only nuclear power plants intended for civilian use are included in this Special Report. Reactor lifetimes are assumed to be 35 years for all light water reactors and 30 years for all other reactor types, unless other data or definite decommissioning dates have been published by the operators. (orig./UA)

  12. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    The listing indicates percentage ownership of commercial nuclear power plants by utility companies as of August 15, 1977. The list includes all plants operating, under construction, docketed for NRC safety and environmental reviews, or under NRC antitrust review. It does not include those plants announced but not yet under review or those plants formally cancelled. In many cases, ownership may be in the process of changing as a result of antitrust license conditions and hearings, altered financial conditions, changed power needs, and other reasons. However, this list reflects only those ownership percentages of which the NRC has been formally notified. Part I lists plants alphabetically with their associated applicants and percentage ownership. Part II lists applicants alphabetically with their associated plants and percentage ownership. Part I also indicates which plants have received operating licenses

  13. Indicators for management of planned outages in nuclear power plants

    International Nuclear Information System (INIS)

    The outages considered within the scope of this publication are planned refuelling outages (PWR and BWR nuclear power plants) and planned outages associated with major maintenance, tests and inspections (PHWR and LWGR nuclear power plants). The IAEA has published some valuable reports providing guidance and assistance to operating organizations on outage management. This TECDOC outlines main issues to be considered in outage performance monitoring and provides guidance to operating organizations for the development and implementation of outage programmes which could enhance plant safety, reliability and economics. It also complements the series of reports published by the IAEA on outage management and on previous work related to performance indicators developed for monitoring different areas of plant operation, such as safety, production, reliability and economics. This publication is based upon the information presented at a technical meeting to develop a standardized set of outage indicators for outage optimization, which was organised in Vienna, 6-9 October 2003. At this meeting, case studies and good practices relating to performance indicator utilization in the process of planned outage management were presented and discussed

  14. Upgrading of fire safety in Indian nuclear power plants

    International Nuclear Information System (INIS)

    Indian nuclear power programme started with the installation of 2 nos. of Boiling Water Reactor (BWR) at Tarapur (TAPS I and II) of 210 MWe each commissioned in the year 1996. The Pressurized Heavy Water Reactor (PHWR) programme in the country started with the installation of 2x220 MWe stations at Rawatbhatta near Kota (RAPS I and II) in the State of Rajasthan in the sixties. At the present moment, the country has 10 stations in operation. Construction is going on for 4 more units of 220 MWe where as work on two more 500 MWe units is going to start soon. Fire safety systems for the earlier units were engineered as per the state-of-art knowledge available then. The need for review of fire protection systems in the Indian nuclear power plants has also been felt since long almost after Brown's Ferry fire in 1975 itself. Task forces consisting of fire experts, systems design engineers, O and M personnel as well as the Fire Protection engineers at the plant were constituted for each plant to review the existing fire safety provisions in details and highlight the upgradation needed for meeting the latest requirements as per the national as well as international practices. The recommendations made by three such task forces for the three plants are proposed to be reviewed in this paper. The paper also highlights the recommendations to be implemented immediately as well as on long-term basis over a period of time

  15. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    The following list indicates percentage ownership of commercial nuclear power plants by utility companies as of December 1, 1979. The list includes all plants licensed to operate, under construction, docketed for NRC safety and envionmental reviews, or under NRC antitrust review. It does not include those plants announced but not yet under review or those plants formally cancelled. In many cases, ownership may be in the process of changing as a result of antitrust license conditions and hearings, altered financial conditions, changed power needs, and other reasons. However, this list reflects only those ownership percentages of which the NRC has been formally notified

  16. Nuclear power plant

    International Nuclear Information System (INIS)

    The present invention provides reduction of volumes of buildings and the area of the buildings having a plurality of plants. Namely, a plurality of reactor building areas corresponding to a plurality of plants are disposed in a reactor building. In each of the reactor building areas, apparatuses including reactors of the plants are disposed. In addition, a plurality of reactor wells, spent fuel pools, pools for temporary storing facilities are arranged on an identical line. With such a constitution, since various kinds of facilities in the reactor buildings can be used in common, the volume of buildings and areas for the buildings in the plants as a whole can be reduced. In addition, turbine buildings can be used in common by the same idea. As a result, the number of buildings, the volume of the buildings and the area of the buildings of the plants can be reduced as a whole thereby enabling to reduce the steps, costs and materials required for building construction. (I.S.)

  17. Nuclear power plant

    International Nuclear Information System (INIS)

    The nuclear part with the negative pressure control system is installed in an underground chamber of a mountain. The containment consists of a sealing concrete layer directly sprayed to the rock and containing reinforcement inserts as well as of a consolidating concrete shell. The sealing concrete layer is combined with the rock by means of prestressed concrete tie rods. (DG)

  18. BWR-X experimental reactor for operating in remote sites (thermal-hydraulic study)

    International Nuclear Information System (INIS)

    The primary objective of this indigenous nuclear plant is to power supply no ease accessible places needed of support, like scientific bases and government facilities. What's more, design and built this particular type BWR at low pressure and temperature will give substantial background on nuclear engineering. The paper is intended to treat the nuclear reactor as a heat source in a power cycle and to study the processes by which energy is removed and converted to useful power. (author)

  19. APROS nuclear plant analyser

    International Nuclear Information System (INIS)

    The paper describes the build-up of the Loviisa plant primary circuit model using graphical user interface and generic components. The secondary circuit model of Loviisa is constructed in the same manner. The entire power plant model thus obtained is used for the calculation of two example transients. These examples originate from the Loviisa 2 unit dynamical tests in 1980. The Modular Plant Analyser results are compared with the Loviisa Unit 2 measurement data. This comparison indicates good agreement with the data. The present work has been performed using the Alliant FX/40 minisupercomputer. With this computer the Loviisa model fulfills at present the real-time requirement with 0.5 second timestep. (orig./DG)

  20. Nuclear reactor plant

    International Nuclear Information System (INIS)

    The plant consists mainly of a steam-raising unit and a steam turbine with high pressure and low pressure stages. There is at least one superheater or intermediate superheater between the steam-raising unit and the low pressure stage. In order to improve the plant efficiency, a high temperature reactor is provided as a source of heat for the superheater or intermediate superheater, which supplies the superheater heat with an energy efficiency of over 60%. This increases the overall net efficiency from 33% to over 36%. (orig.)

  1. Nuclear plant undergrounding

    International Nuclear Information System (INIS)

    Under Section 25524.3 of the Public Resources Code, the California Energy Resources Conservation and Development Commission (CERCDC) was directed to study ''the necessity for '' and the effectiveness and economic feasibility of undergrounding and berm containment of nuclear reactors. The author discusses the basis for the study, the Sargent and Lundy (S and L) involvement in the study, and the final conclusions reached by S and L

  2. Submarine nuclear power plant

    International Nuclear Information System (INIS)

    Purpose: To provide a ballast tank, and nuclear power facilities within the containment shell of a pressure resistance structure and a maintenance operator's entrance and a transmission cable cut-off device at the outer part of the containment shell, whereby after the construction, the shell is towed, and installed by self-submerging, and it can be refloated for repairs by its own strength. Constitution: Within a containment shell having a ballast tank and a pressure resisting structure, there are provided nuclear power facilities including a nuclear power generating chamber, a maintenance operator's living room and the like. Furthermore, a maintenance operator's entrance and exit device and a transmission cable cut-off device are provided within the shell, whereby when it is towed to a predetermined a area after the construction, it submerges by its own strength and when any repair inspection is necessary, it can float up by its own strength, and can be towed to a repair dock or the like. (Yoshihara, H.)

  3. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  4. Nuclear power plant with pressure vessel boiling water reactor VK-300 for district heating and electricity supply

    International Nuclear Information System (INIS)

    The viability for Russia of the Boiling Water Reactor (BWR) concept has been shown by a number of feasibility studies fulfilled for perspective sites with increased energy demands. Russia has long (31 year) successful experience in operation of NPPs with the vessel-type boiling reactor VK-50 which is located in the city of Dimitrovgrad. Taking into account the large Russian district heating market, it is expedient to apply this concept (BWR) not only for electricity supply, but also for district heating. This is a way to increase of nuclear power plant competitiveness along with good safety performance. The safety and protection of nuclear heat customer is guaranteed by reliable technical means which are well checked at Russian nuclear sites. (author)

  5. Alternative BWR plant shutdown system using fire protection system and containment vent

    International Nuclear Information System (INIS)

    In certain severe accidents, caused by factors beyond the plant design bases, a loss of operability in the systems normally used to attain safe shutdown of the plant could occur. For such situations, Cofrentes NPP has carried out a series of studied and modifications that aim to attain safe shutdown using alternative systems which have their own power supply systems and are not dependent on external or emergency networks. These alternative systems include some existing plant systems (fire protection system, FPS) which have been validated to perform new operations, as well as other systems wich have been developed recently (containment venting system, CVS). Reactor cooling is restored through the injection of water using the FPS system (diesel pump), and the residual heat removal function is performed by discharging water through the relief valves into the suppression pool and from there into the external atmosphere by means of the CVS. (Author) 6 refs

  6. Optimization and improvement of the technical specifications for Santa Maria de Garona and Cofrentes nuclear power plants

    International Nuclear Information System (INIS)

    Technical Specifications (TS) form one of the basic documents necessary for licensing nuclear power plants and are required by the Government in accordance with Article 26 of the Regulation for Nuclear and Radioactive Facilities. They contain specific plant characteristics and operating limits to provide adequate protection for the safety and health of operators and the general public. For operator actuation, TS include all the surveillance requirements and limiting operating conditions (operation at full power, startup, hot and cold shutdown, and refueling outage) of safety-related systems. They also include the conventional support systems which are necessary to keep the plant in a safe operating conditioner to bring it to safe shutdown in the event of incidents or hypothetical accidents. Because of the large volume of information contained in the TS, the NRC and American utility owners began to simplify and improve the initial standard TS, which has given way to the development of a TS Optimization Program in the USA under the auspices of the NRC. Empresarios Agrupados has been contracted by the BWR Spanish Owners' Group (GPE-BWR) to develop optimized TS for the Santa Maria de Garona and Cofrentes Nuclear Power Plants. The optimized and improved TS are simplified versions of the current ones and facilitate the work of plant operators. They help to prevent risks, and reduce the number of potential transients caused by the large number of tests required by current TS. Plant operational safety is enhanced and higher effective operation is achieved. The GPE-BWR has submitted the first part of the optimized TS with their corresponding Bases to the Spanish Nuclear Council (CSN), for comment and subsequent approval. Once the TS are approved by the Spanish Nuclear Council, the operators of the Santa Maria de Garona and Cofrentes Nuclear Power Plants will be given a training and adaptation course prior to their implementation. (author)

  7. Risk informed plant modifications and upgrades for life extension of Indian BWR

    International Nuclear Information System (INIS)

    Probabilistic Safety Assessment techniques can be used to determine how ageing is affecting component and system unavailability; to identify plant components and systems that have a propensity to ageing and are risk significant. An essential requirement for PSA is the failure data for various components. An analysis of the failure data TAPS operation revealed that they were still in their useful life period. Hence, age related component modelling were not called for. With the objective of TAPS life extension, NPCIL initiated Level - 1 Probabilistic Safety Assessment (PSA) studies for Tarapur Atomic Power Station - 1 and 2. This Plant specific PSA Level - 1 study resulted in identifying and understanding the key plant vulnerabilities and in making recommendations for improvements by identifying risk significant systems and component and in particular, in case of TAPS, PSA has proved to be a valuable tool to understand the implications of non - conformance to the current standards and in making decisions on upgrades. The same methodology was used for trying out the different configurations of the system under consideration for upgrade and for optimizing the final decision on the same. These results are also useful for prioritizing the implementation of the upgrades/modifications, technical specification optimisation, where the maximum benefit for plant safety can be obtained. Insights available based on this risk informed study for TAPS life extension are discussed in this paper. (author)

  8. Operating the plant, quality assurance, and the job of the operating staff, Volume Twelve

    International Nuclear Information System (INIS)

    Subject matter includes operating the plant (the role of the operator, the control room, plant technical specifications, plant operating procedures, initial startup program, BWR/PWR plant startup, BWR/PWR steady state power operation, BWR/PWR transient operation, emergency operation), quality assurance (what is quality, what is quality control, quality assurance includes quality control, government regulation and quality assurance, administrative controls for nuclear power plants, the necessity of reviews and audits, practical quality assurance), and the job of the operating staff (the plant operating staff, plant safety, first aid and resuscitation, general plant hazards, personnel protective equipment, handling chemicals, handling compressed gas, equipment repair and maintenance, communicating with others

  9. Marine nuclear power plant simulator

    International Nuclear Information System (INIS)

    In this paper, a real time simulator of nuclear power plant for ship propulsion is described. The simulator is composed of a hybrid computer and an operating console. Simulated power plant is a 36 MWt PWR plant. Reactor kinetics, fuel temperature, primary coolant temperature, primary pressure, temperature and pressure of steam in the steam generator, steam flow, rotating speed of the propeller shaft, and various controlling system are simulated. Using this simulator, frequency response functions of the reactor power to the variation of reactivity and load were measured. (author)

  10. Elecnuc. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    This 2003 version of Elecnuc contents information, data and charts on the nuclear power plants in the world and general information on the national perspectives concerning the electric power industry. The following topics are presented: 2002 highlights; characteristics of main reactor types and on order; map of the French nuclear power plants; the worldwide status of nuclear power plants on 2002/12/3; units distributed by countries; nuclear power plants connected to the Grid by reactor type groups; nuclear power plants under construction; capacity of the nuclear power plants on the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear plants by country at the end 2002; performance indicator of french PWR units; trends of the generation indicator worldwide from 1960 to 2002; 2002 cumulative Load Factor by owners; nuclear power plants connected to the grid by countries; status of license renewal applications in Usa; nuclear power plants under construction; Shutdown nuclear power plants; exported nuclear power plants by type; exported nuclear power plants by countries; nuclear power plants under construction or order; steam generator replacements; recycling of Plutonium in LWR; projects of MOX fuel use in reactors; electricity needs of Germany, Belgium, Spain, Finland, United Kingdom; electricity indicators of the five countries. (A.L.B.)

  11. Nuclear power plant

    International Nuclear Information System (INIS)

    The plant of the present invention can suppress the amount of clad in feedwater when drains of a moisture content separation heater or a high pressure feedwater heater are recovered. That is, the moisture content separation heater has ferrite or austenite type stainless steel heat transfer pipes. A chromium-enriched layer is formed on the surface of the heat transfer pipe by chromizing treatment or flame spraying. Then, a stainless steel heat transfer pipe having chromium-enriched layer is incorporated to at least one of the moisture content separation heater or the high pressure feedwater heater. During plant operation, the temperature of heated steams is as high as 235 to 282degC. Accordingly, this is a severe corrosion region for ferrite or austenite stainless steel. However, the chromium-enriched layer of excellent corrosion resistance is formed on the surface of the heat transfer pipe. Accordingly, metal ingredients are less leached. As a result, even if the drains are recovered to feedwater, increase of concentration of the clads in the feedwater can be prevented. (I.S.)

  12. Small, moveable nuclear power plant

    International Nuclear Information System (INIS)

    Most reliable projections indicate that by the year 2000 approximately 25% of the world's electrical energy and possibly 10% of its process heat for industry will be provided by nuclear power plants. Whilst much of this will come from large nuclear steam raising units-in excess of 900 M.W.(e)-an increasing proportion will be provided from small modular plant, designed for portabiity or mobility. The technical, environmental and economic parameters of such a plant which will usually be of a size around 600 M.W.(e) or less and which in general terms derives its basic design features from nuclear marine propulsion units is reviewed. If suitably standardised, such a nuclear plant, mounted on barges, installed in ships or even airlifted into difficult locations, could provide an attractive solution to the energy problems of developing countries or remote communities. Apart from electrical energy production, the units could operate in a duplex mode to supply steam for central heating, sea-water or bore water distillation or industrial process heat. (Author)

  13. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  14. Abundance and distribution of radionuclides discharged from a BWR nuclear power station into a marine bay

    International Nuclear Information System (INIS)

    This article summarizes a portion of one of a series of radiological surveillance studies conducted by the U.S. Environmental Protection Agency (EPA) at nuclear power stations. Radionuclide concentrations were measured in Barnegat Bay at the New Jersey shore during EPA's 2-year radiological surveillance study at the Oyster Creek Nuclear Generating Station. The station discharges batches of radioactive liquid waste into effluent condenser cooling water, which flows through Oyster Creek into Barnegat Bay 3 km from the point of discharge. The bay is long, narrow, and shallow, with few passages to the Atlantic Ocean. Radionuclide concentrations were measured repeatedly in water, sediment, marine vegetation, fish, clams, and crabs at various sampling points. Measured values were compared to calculated values based on bioaccumulation factors, and both sets were used to compute dose equivalent rates to the most exposed persons in the environment. The two observed critical radiation exposure pathways - fish consumption and standing on beaches - resulted in dose equivalents of less than 1 mrem/year; the critical radionucides were 90Sr and 60Co, respectively. Indicator radionuclides were identified, and environmental radiological monitoring activities were recommended

  15. Elecnuc. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    This small booklet summarizes in tables all data relative to the nuclear power plants worldwide. These data come from the IAEA's PRIS and AREVA-CEA's GAIA databases. The following aspects are reviewed: 2007 highlights; Main characteristics of reactor types; Map of the French nuclear power plants on 2007/01/01; Worldwide status of nuclear power plants (12/31/2007); Units distributed by countries; Nuclear power plants connected to the Grid- by reactor type groups; Nuclear power plants under construction on 2007; Evolution of nuclear power plants capacities connected to the grid; First electric generations supplied by a nuclear unit in each country; Electrical generation from nuclear power plants by country at the end 2007; Performance indicator of French PWR units; Evolution of the generation indicators worldwide by type; Nuclear operator ranking according to their installed capacity; Units connected to the grid by countries at 12/31/2007; Status of licence renewal applications in USA; Nuclear power plants under construction at 12/31/2007; Shutdown reactors; Exported nuclear capacity in net MWe; Exported and national nuclear capacity connected to the grid; Exported nuclear power plants under construction; Exported and national nuclear capacity under construction; Nuclear power plants ordered at 12/31/2007; Long term shutdown units at 12/31/2007; COL (combined licences) applications in the USA; Recycling of Plutonium in reactors and experiences; Mox licence plants projects; Appendix - historical development; Meaning of the used acronyms; Glossary

  16. Operator support architecture for monitoring abnormal symptoms of nuclear power plant based on knowledge engineering

    International Nuclear Information System (INIS)

    An architecture to support nuclear power plant operators for monitoring abnormal symptoms has been proposed based on the techniques of knowledge engineering, and the feasibility of a plant monitoring support system was investigated. The purpose of the support system is to present the operators with useful information so that they can make correct judgment at an early and subtle stage of abnormal plant conditions. In the architecture proposed, abductive reasoning is performed to search for causal events and deductive one to predict consequential events using the knowledge representing plant components as frames and those representing causal relations as production rules. A method to deal with uncertainties in each types of reasoning has been adopted, and it is used to rank several hypotheses of causal events and to assess the importance of plant parameters for monitoring. A prototype system was developed, and its usefulness was tested using a case of failure in a recirculation pump of a BWR plant. (author)

  17. Regression analysis of technical parameters affecting nuclear power plant performances

    International Nuclear Information System (INIS)

    Since the 80's many studies have been conducted in order to explicate good and bad performances of commercial nuclear power plants (NPPs), but yet no defined correlation has been found out to be totally representative of plant operational experience. In early works, data availability and the number of operating power stations were both limited; therefore, results showed that specific technical characteristics of NPPs were supposed to be the main causal factors for successful plant operation. Although these aspects keep on assuming a significant role, later studies and observations showed that other factors concerning management and organization of the plant could instead be predominant comparing utilities operational and economic results. Utility quality, in a word, can be used to summarize all the managerial and operational aspects that seem to be effective in determining plant performance. In this paper operational data of a consistent sample of commercial nuclear power stations, out of the total 433 operating NPPs, are analyzed, mainly focusing on the last decade operational experience. The sample consists of PWR and BWR technology, operated by utilities located in different countries, including U.S. (Japan)) (France)) (Germany)) and Finland. Multivariate regression is performed using Unit Capability Factor (UCF) as the dependent variable; this factor reflects indeed the effectiveness of plant programs and practices in maximizing the available electrical generation and consequently provides an overall indication of how well plants are operated and maintained. Aspects that may not be real causal factors but which can have a consistent impact on the UCF, as technology design, supplier, size and age, are included in the analysis as independent variables. (authors)

  18. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  19. Passive Nuclear Plants Program (UPDATE)

    International Nuclear Information System (INIS)

    The light water passive plants program (PCNP), today Advanced Nuclear Power Plants Program (PCNA), was constituted in order to reach the goals of the Spanish Electrical Sector in the field of advanced nuclear power plants, optimize the efforts of all Spanish initiatives, and increase joint presence in international projects. The last update of this program, featured in revision 5th of the Program Report, reflects the consolidation of the Spanish sector's presence in International programs of the advanced power plants on the basis of the practically concluded American ALWR program. Since the beginning of the program , the PCNP relies on financing from the Electrical sector, Ocide, SEPI-Endesa, Westinghouse, General Electric, as well as from the industrial cooperators, Initec, UTE (Initec- Empresarios Agrupados), Ciemat, Enusa, Ensa and Tecnatom. The program is made up of the following projects, already concluded: - EPRI's Advanced Light Water Plants Certification Project - Westinghouse's AP600 Project - General Electric's SBWR Project (presently paralyzed) and ABWR project Currently, the following project are under development, at different degrees of advance: - EPP project (European Passive Plant) - EBWR project (European Advanced Boiling Water Reactor)

  20. ALMOS-2 - computer program for accident analyses of BWR-type reactor plants. Pt. 1

    International Nuclear Information System (INIS)

    ALMOS-2 is a non-linear, one-dimensional digital computer program for accident analyses of boiling water reactor plants. The program is capable of calculating transients with large amplitude deviations from steady state conditions, such as pump failures, loss of heat sink, feedwater system failures, control rod malfunction and safety system failures. The program is a suitable tool for the analysis of Anticipated Transients Without Scram (ATWS). The model includes the reactor core, the main coolant system and the safety system. Boundary conditions to these systems can be defined as functions of time by input data. The program description includes a short introduction into the physical model, a presentation of the program structure and program components and a detailed input data description. For test purposes the results of a transient calculation are added. (orig.)

  1. Safety for nuclear power plants

    International Nuclear Information System (INIS)

    The September 11, 2001 terrorist attacks on the World Trade Center in New York have reopened the debate in Germany abut the safety of existing nuclear power plants. On the one hand, it culminated in the demand, made from a prominent political position, to disconnect from the grid prematurely five of the older nuclear power plants because they offered no sufficient protection from terrorist attacks carried out by means of commercial airliners. On the other hand, the competent German Federal Ministry of the Interior to this day has not been able to detect a hazardous situation for Germany which would require nuclear power plants (or other facilities) to be shut down - temporarily -so as to reduce their hazard potential. The legal situation is discussed in detail. It is evident that the debate about the safety of any kind of industrial plants will go on and will be revived by any kind of event. Despite the growing need for public safety, it must not be overlooked in the public debate that an absolute exclusion of risk is impossible and that, consequently, there can be no absolute protection of any legal object. (orig.)

  2. Docommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    The German utilities operating nuclear power plants have long concerned themselves with aspects of decommissioning and for this purpose an engineering company was given a contract to study the entire spectrum of decommissioning. The results of this study have been available in autumn 1980 and it is possible to discuss all the aspects of decommissioning on a new basis. Following these results no change in the design concept of LWR nuclear power plants in operation or under construction is necessary because the techniques, necessary for decommissioning, are fully available today. The technical feasibility of decommissioning for power plants of Biblis A and KRB type has been shown in detail. The calculations of the quantity of waste produced during removal of a nuclear power plant could be confirmed and it could be determined with high procedure. The radiation dose to the decommissioning personnel is in the range of the radiation protection regulations and is in the same range as the radiation dose to the personnel within a yearly inservice inspection. (AF)

  3. Fire prevention in nuclear plants

    International Nuclear Information System (INIS)

    About 80 fire starts are reported in EDF nuclear power plants every year but only 3 or 4 turn into a real fire and none has, so far, has led to a major safety failure of a nuclear plant. A new regulation has been implemented in july 2014 that strengthens the concept of defense in depth, proposes an approach that is proportionate to the stakes and risks, this proportionality means that the requirements for a power reactor are not the same as for a nuclear laboratory, and imposes an obligation or result rather than of means. The second article deals with the fire that broke out in the waste silo number 130 at La Hague plant in january 1981. The investigation showed that the flammability of the silo content had been underestimated. The third article presents the consequences of the fire that broke out in a power transformer at the Cattenom plant in june 2013. The fire was rapidly brought under control thanks to the immediate triggering of the emergency plan. The article details also the feedback experience of this event. (A.C.)

  4. Effectiveness of a large mimic panel in an existing nuclear power plant central control board

    International Nuclear Information System (INIS)

    We conducted the analysis of the nuclear power plant (NPP) operators' behaviors under emergency conditions by using training simulators as a joint research project by Japanese BWR groups for twelve years. In the phase-IV of this project we executed two kinds of experiments to evaluate the effectiveness of the interfaces. One was for evaluations of the interfaces such as CRTs with touch screen, a large mimic panel, and a hierarchical annunciator system introduced in the newly developed ABWR type central control board. The other was that we analyzed the operators' behaviors in emergency conditions by using the first generation BWR type central control board which was added new interfaces such as a large display screen and demarcation on the board to help operators to understand the plant. The demarcation is one of the visual interface improvements and its technique is that a line enclosing several components causes them to be perceived as a group.The result showed that both the large display panel Introduced in ABWR central control board and the large display screen in the existing BWR type central control board improved the performance of the NPP operators in the experiments. It was expected that introduction of the large mimic panel into the existing BWR type central control boards would improve operators' performance. However, in the case of actual installation of the large display board into the existing central control boards, there are spatial and hardware constraints. Therefore the size of lamps, lines connecting from symbols of the pumps or valves to the others' will have to be modified under these constraints. It is important to evaluate the displayed information on the large display board before actual installation. We made experiments to solve these problems by using TEPCO's research simulator which is added a large mimic panel. (author)

  5. Accumulation of operator workload data by using A-BWR training simulator

    International Nuclear Information System (INIS)

    Human-machine interface (HMI) of A-BWR has been developed in order to improve operational safety, reliability and to reduce workload. A-BWR HMI is fully computerized. JNES (Japan Nuclear Energy Safety Organization) and BTC (BWR Operator Training Center Corporation) have accumulated the operator workload quantitative data, related to the observation and operation at typical transient conditions, in order to evaluate the difference of operational workloads between A-BWR HMI and conventional type HMI. The workload evaluation shows the following results: - The workload density (observation and operation frequencies per unit time) of ABWR just after the plant trip is less than that of BWR-5. - At stable conditions after the transient, the workload density of ABWR becomes higher comparing that of BWR-5. A-BWR alarm system may increase the workload density caused by alarm multi-layer structure, because an operator has to use the flat and/or the CRT display to pursue every alarm. The analysis of shift team training at BTC shows that total workload is reduced at ABWR but alarm confirmation work still remains as burden. These results show some modifications might be needed for future HMI. To grasp the tendency of operator workload difference by the control panel type difference, the operator workload quantitative data have accumulated using ABWR type simulator and conventional type simulator at the same typical transient condition. These data were arranged operation frequency data, number of alarm generating data and number of switching CRT pictures data according to plant behaviour. The ABWR type HMI's characteristic have become clear by these operator workload data and the team characteristic evaluation data which BTC evaluated comparing the team performance difference of HMI type

  6. Atucha I nuclear power plant transients analysis

    International Nuclear Information System (INIS)

    A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

  7. Providing emergency supply of nuclear power plants

    OpenAIRE

    ROZMILER, Jiří

    2013-01-01

    Work "Providing emergency power nuclear power plant" describes how solving their own consumption nuclear power plant, as emergency power supply is designed and how it should be a solution of known states of emergency, having an immediate impact on the power consumption of their own nuclear power plants. The aim of this thesis is to propose options to strengthen its own emergency power consumption of nuclear power plants, one might say-more resistant to harsh extremes, which could lead to loss...

  8. 60Co deposition on carbon-steel structural materials after seawater infiltration in BWR plant

    International Nuclear Information System (INIS)

    Seawater infiltration occurred during shutdown of the Hamaoka Unit 5 (H-5). Chloride ion (Cl-) is known to affect the corrosion behavior of carbon steel, and it may change the properties of the oxide film formed on the surface. 60Co deposition in high-temperature water is strongly related to the oxide film properties, and any change in the properties may affect the 60Co deposition after the plant is restarted. This paper shows the results of 60Co deposition tests of carbon steels under simulated H-5 water conditions. Specimens for the 60Co deposition tests were prepared in three steps, which simulated the conditions of normal plant operation, seawater infiltration, and chemical decontamination after the infiltration. The first step was a prefilming step under Normal Water Condition (NWC). The second step included two different conditions: seawater infiltration and keeping after infiltration. Prefilmed specimens were immersed in 450 ppm Cl- diluted artificial seawater at 513 K for 24 hours. Following that, the specimens were immersed in 50 ppm Cl- diluted artificial seawater at 323 K for 100-500 hours. During the second step, the prefilming oxide (NiFe2O4) flaked off in spots. In the third step, the oxide remaining on some specimens after the second step was removed chemically. The three types of prepared specimens, that is, a prefilmed specimen, an exposed specimen, and an oxide-removed specimen, were used for the 60Co deposition tests using 0.015 Bq/cm3 60Co solution for 500 or 1000 hours under NWC conditions. After the deposition tests, the 60Co activity was measured with a Ge detector. From the results of the deposition test, at the spots where flaking occurred in the second step, only loose hematite was formed, and generation of a new protective film was not observed. The amount of 60Co deposited on the exposed specimen was more than that on the prefilmed and oxide-removed specimens. The simulated infiltrating conditions inhibited the regeneration of a

  9. Trial application of guidelines for nuclear plant response to an earthquake

    International Nuclear Information System (INIS)

    Guidelines have been developed to assist nuclear plant personnel in the preparation of earthquake response procedures for nuclear power plants. These guidelines are published in EPRI report NP-6695, ''Guidelines for Nuclear Plant Response to an Earthquake,'' dated December 1989. This report includes two sets of nuclear plant procedures which were prepared to implement the guidelines of EPRI report NP-6695. The first set were developed by the Toledo Edison Company Davis-Besse plant. Davis-Besse is a pressurized water reactor (PWR) and contains relatively standard seismic monitoring instrumentation typical of many domestic nuclear plants. The second set of procedures were prepared by Yankee Atomic Electric Company for the Vermont Yankee facility. This plant is a boiling water reactor (BWR) with state-of-the-art seismic monitoring and PC-based data processing equipment, software developed specifically to implement the OBE Exceedance Criterion presented in EPRI report NP-5930, ''A Criterion for Determining Exceedance of the operating Basis Earthquake.'' The two sets of procedures are intended to demonstrate how two different nuclear utilities have interpreted and applied the EPRI guidance given in report NP-6695

  10. Outlet device to limit excess pressure in nuclear power plants, in particular BWR reactor plants

    International Nuclear Information System (INIS)

    The outlet flow side of a condensation pipe immersed into a water reservoir is equipped with internals. These are a concentric orifice tube with an annular gap, a perforated hollow cone, or a multitude of annular diaphragms. The annular diaphragms have sections of perforated orifices. The internals can be used to initiate a spray action at the overflow edges multiplying the heat transfer area of the cooling water relative to the steam flow. (DG)

  11. Mechanized ultrasonic examination of piping systems in nuclear power plants

    International Nuclear Information System (INIS)

    The success of mechanized ultrasonic examination applied on welds in piping systems in nuclear power plants is highly dependent on its careful preparation. From the development of an adequate examination technique to its implementation on site, many problems are to be solved. This is especially the case when dealing with austenitic welds or dissimilar metal welds. In addition to the specific needs for examination technique based on material properties and requirements for minimum flaw size detection, accessibility and radiation aspects have to be considered. A crew of skilled and highly trained examination personnel is required. Experience in various nuclear power plants, - BWR's and PWR's of different designs - has shown, that even difficult examination problems can be successfully solved, provided that there is a good preparation. The necessary step by step proceeding is illustrated by examples concerning mechanized examination. Preservice inspections and in-service inspections with specific requirements, due to the types of flaws to be found or the type of material concerned, are discussed

  12. Design for reactor core safety in nuclear power plants

    International Nuclear Information System (INIS)

    This Guide covers the neutronic, thermal, hydraulic, mechanical, chemical and irradiation considerations important to the safe design of a nuclear reactor core. The Guide applies to the types of thermal neutron reactor power plants that are now in common use and fuelled with oxide fuels: advanced gas cooled reactor (AGR), boiling water reactor (BWR), pressurized heavy water reactor (PHWR) (pressure tube and pressure vessel type) and pressurized water reactor (PWR). It deals with the individual components and systems that make up the core and associated equipment and with design provisions for the safe operation of the core and safe handling of the fuel and other core components. The Guide discusses the reactor vessel internals and the reactivity control and shutdown devices mounted on the vessel. Possible effects on requirements for the reactor coolant, the reactor coolant system and its pressure boundary (including the pressure vessel) are considered only as far as necessary to clarify the interface with the Safety Guide on Reactor Coolant and Associated Systems in Nuclear Power Plants (IAEA Safety Series No. 50-SG-D13) and other Guides. In relation to instrumentation and control systems the guidance is mainly limited to functional requirements

  13. Testing programs related to potential adverse flow effects in nuclear power plants

    International Nuclear Information System (INIS)

    Potential adverse flow effects on nuclear power plant Structures, Systems and Components (SSCs), can result from various causes. For example, reactor vessel and main steam system piping and components (including the steam dryer, safety relief and power-operated valves, and pipe supports) in BWR nuclear power plants can be damaged by pressure fluctuations and vibration resulting from acoustic resonances occurring in the main steam system or reactor vessel. The acoustic resonance phenomenon can also occur in pressurized water reactor (PWR) nuclear power plants with resulting damage to plant piping and components. Sampling probes in feedwater and condensate systems in nuclear power plants are also susceptible to adverse flow effects. The NRC staff reviews the evaluation performed by applicants for the construction and operation of nuclear power plants under 10 CFR Part 50 or Part 52 of the potential for adverse flow effects on plant SSCs. This proposed paper will discuss the various aspects of the programs established by applicants and licensees for monitoring plant data to verify that adverse flow effects are not occurring. These aspects of the test programs include the following: conducting walk-downs, and inspecting components during power ascension and operation at full licensed power conditions, dynamic response analysis of reactor internals under operational flow transients and steady-state conditions; preoperational flow-induced vibration testing of reactor internals; correlation of the test results with the analytical results. (authors)

  14. QA programs in nuclear power plants

    International Nuclear Information System (INIS)

    As an overview of quality assurance programs in nuclear power plants, the energy picture as it appears today is reviewed. Nuclear power plants and their operations are described and an attempt is made to place in proper perspective the alleged ''threats'' inherent in nuclear power. Finally, the quality assurance programs being used in the nuclear industry are described

  15. Contingency programs for BWR pipe cracking

    International Nuclear Information System (INIS)

    General Electric (GE) has aggressively addressed the problem of Intergranular Stress Corrosion Cracking (IGSCC). Intergranular Stress Corrosion Cracking has occurred in boiling water reactors in less than 1% of the Type-304 stainless steel welds in operating plants. However a comprehensive program by GE, with EPRI support in many cases, has provided technical solutions to limit IGSCC. As part of this program the Nuclear Services Department (NSD) has taken new technology and applied it to programs for the prevention of IGSCC and for repair of pipe cracks should they occur. The purpose of this paper is to provide a description and the current status of the major ongoing service programs for dealing with the problems related to IGSCC in BWR operating plants

  16. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Expert robots will enhance safety and operations in nuclear plants. E. I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligent, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  17. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Expert robots enhance a safety and operations in nuclear plants. E.I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligence, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  18. Elecnuc. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1997 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1997; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; forecasts; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

  19. Nuclear power plants and environment

    International Nuclear Information System (INIS)

    The question of nuclear power plants is analysed in details. The fundamental principles of reactors are described as well as the problems of safety involved with the reactor operation and the quantity and type of radioactive released to the environment. It shows that the amount of radioactive is very long. The reactor accidents has occurred, as three mile island, are also analysed. (M.I.A.)

  20. Investigation of valve failure problems in LWR power plants

    International Nuclear Information System (INIS)

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

  1. Financing nuclear power plant decommissioning

    International Nuclear Information System (INIS)

    Much is at stake in developing a financial strategy for decommissioning nuclear power plants. Since decommissioning experience is limited to relatively small reactors, will the costs associated with larger reactors be significantly higher. Certainly the decommissioning issue intersects with other critical issues that will help to determine the future of commercial nuclear power in the US. The author examines briefly the basic concepts and terms related to decommissioning expenses, namely: (1) segregated fund; (2) non-segregated fund; (3) external method; and (4) internal method. He concludes that state regulatory commissions have turned increasingly to the external funding method because of increasing costs and related problems associated with nuclear power, changing conditions and uncertainties concerned with utility restructuring, and recent changes in federal tax laws related to decommissioning. Further, this trend is likely to continue if financial assurance remains a primary concern of regulators to protect this public interest

  2. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  3. Seismic resistance design of nuclear power plant building structures in Japan

    International Nuclear Information System (INIS)

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  4. Study of the optimization of maintenance plan for nuclear power plants

    International Nuclear Information System (INIS)

    This paper proposes a quantitative evaluation method for the maintenance plan for nuclear power plants, developed by introducing the scientific approach, and also proposes a method to search for an optimum maintenance plan to be obtained by maximizing nuclear safety and economic efficiency simultaneously, then balancing them. As a result of consideration, the following results were obtained. (1) The quantitative evaluation methodology for optimizing the maintenance plan for nuclear power plants was developed. (2) The computer simulation of maintenance planning for a couple of BWR systems by using this methodology was carried out. It was concluded that this methodology can produce a new maintenance plan which meets the maintenance targets corresponding to optimum maintenance. (author)

  5. Start up and commercial operation of Laguna Verde nuclear power plant. Unit 1

    International Nuclear Information System (INIS)

    Prior to start up of Laguna Verde nuclear power plant preoperational tests and start tests were performed and they are described in its more eminent aspects. In relation to commercial operation of nuclear station a series of indicator were set to which allow the measurement of performance in unit 1, in areas of plant efficiency and personal safety. Antecedents. Laguna Verde station is located in Alto Lucero municipality in Veracruz state, 70 kilometers north-northeast from port of Veracruz and a 290 kilometers east-northeast from Mexico city. The station consist of two units manufactured by General Electric, with a nuclear system of vapor supply also called boiling water (BWR/5), and with a system turbine-generator manufactured by Mitsubishi. Each unit has a nominal power of 1931 MWt and a level design power of 675 Mwe and a net power of 654 Electric Megawatts

  6. Sabotage at Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Purvis, James W.

    1999-07-21

    Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented.

  7. Sabotage at Nuclear Power Plants

    International Nuclear Information System (INIS)

    Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented

  8. Occupational dose control in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Reduction in occupational exposure at nuclear power plants is desirable not only in the interest of the health and safety of plant personnel, but also because it enhances the safety and reliability of the plants. This report summarises the current trends of doses to workers at nuclear power plants and the achievements and developments regarding methods for their reduction

  9. Nuclear plant repowered as cogeneration facility

    International Nuclear Information System (INIS)

    This article describes the first conversion of an idled nuclear plant to natural gas. The new investment was less than building a new gas combined-cycle plant. The nuclear plant was never completed due to financial problems resulting from regulatory and legal issues. The plant was converted as a joint venture between Dow Chemical, Consumers Power, and five other multinational partners

  10. An analysis of human maintenance failures of a nuclear power plant

    International Nuclear Information System (INIS)

    In the report, a study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components appeared to be especially prone to human failures. Many human failures were found in safety related systems. Several failures also remained latent from outages to power operation. However, the safety significance of failures was generally small. Modifications were an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more in the future. (orig.)

  11. Statistical analysis of human maintenance failures of a nuclear power plant

    International Nuclear Information System (INIS)

    In this paper, a statistical study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components are especially prone to human failures. Many human failures were found in safety related systems. Similarly, several failures remained latent from outages to power operation. The safety significance was generally small. Modifications are an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more, in future. (orig.)

  12. BWR Water Chemistry Guidelines: 1993 Revision, Normal and hydrogen water chemistry

    International Nuclear Information System (INIS)

    The goal of water chemistry control is to extend the operating life of the reactor and rector coolant system, balance-of-plant components, and turbines while simultaneously controlling costs to safeguard the continued economic viability of the nuclear power generation investment. To further this goal an industry committee of chemistry personnel prepared guidelines to identify the benefits, risks, and costs associated with water chemistry in BWRs and to provide a template for an optimized water chemistry program. This document replaces the BWR Normal Water Chemistry Guidelines - 1986 Revision and the BWR Hydrogen Water Chemistry Guidelines -- 1987 Revision. It expands on the previous guidelines documents by covering the economic implications of BWR water chemistry control

  13. Modelling of nuclear power plant decommissioning financing

    International Nuclear Information System (INIS)

    Costs related to the decommissioning of nuclear power plants create a significant financial burden for nuclear power plant operators. This article discusses the various methodologies employed by selected European countries for financing of the liabilities related to the nuclear power plant decommissioning. The article also presents methodology of allocation of future decommissioning costs to the running costs of nuclear power plant in the form of fee imposed on each megawatt hour generated. The application of the methodology is presented in the form of a case study on a new nuclear power plant with installed capacity 1000 MW. (authors)

  14. BWR Servicing and Refueling Improvement Program: Phase I summary report

    International Nuclear Information System (INIS)

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

  15. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  16. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133{degrees}F) has a 95-percentile uncertainty of 14.4 K (26{degrees}F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175{degrees}F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6{degrees}F).

  17. Extension of integrated neutronic and thermal-hydraulic analysis capabilities of the 'numerical nuclear reactor' software system for BWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Weber, D.; Sofa, T.; Pointer, D.; Tentner, A.; Zhong, Z. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States); Downar, T.; Thomas, J. [Purdue Univ., 1290 Nuclear Engineering Building, West Lafayette, IN 47907-1290 (United States); Lo, S.; Splawski, A. [CD-adapco, 200 Shepherds Bush Road, London W6 7NL (United Kingdom)

    2006-07-01

    The Numerical Nuclear Reactor is a software analysis system based on the integration of high fidelity models of neutronic, thermal-hydraulic and thermo-mechanical phenomena. Originally developed for pressurized water reactors (PWRs), the current version has been extended to treat neutronic and thermal-hydraulic issues in boiling water reactors (BWRs). The neutronic module is an extension of the DeCART whole core neutron transport code, which is capable of generating three-dimensional sub-pin level power distributions with the thermal feedback effect incorporated directly during the whole core calculation. The original cell based modular ray tracing scheme has been extended to an assembly modular ray tracing to address the complex geometry of a BWR. Thermal-hydraulic analyses are performed with computational fluid dynamics (CFD) solutions based on the STAR-CD code which has been extended to treat boiling two-phase flow, along with conjugate heat transfer, for expected BWR conditions. Verification and validation of the neutronics and thermal-hydraulics modules are described. Results of integrated calculations are illustrated. A review of the numerical performance on parallel computing system is also provided. (authors)

  18. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors

    International Nuclear Information System (INIS)

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  19. Nuclear power plant computer replacement

    International Nuclear Information System (INIS)

    Consumers Power Company (CPCo) has recently completed an ambitious project to replace the non-safety-related plant process computer for the Palisades Nuclear Plant. The replacement computer retains the functions of the old plant computer system with minor exceptions. However, the architecture of the new computer system is vastly improved from the old plant computer. The hardware and software components of the new computer system are the state of the art. The old system consisted of five stand-alone computers, each performing their intended function separately. The new architecture employs a number of computing and data acquisition nodes networked together to gather, share, process, present, and store data more efficiently. For cost efficiency, some of the old hardware, such as the Critical Functions Monitoring System multiplexer, has been retained. The new computer system has a significantly higher availability, and lower maintenance cost. It performs certain functions that were not possible with the old system, such as real-time validation of rod position monitored by the Plant Information Processor and the Secondary Position Indicator. While the new computer system is more capable, more reliable, more user friendly, and easier to maintain, taking it to operability status presented CPCo with some new and challenging issues

  20. Nuclear Plant Integrated Outage Management

    International Nuclear Information System (INIS)

    This paper is a discussion of an emerging concept for improving nuclear plant outage performance - integrated outage management. The paper begins with an explanation of what the concept encompasses, including a scope definition of the service and descriptions of the organization structure, various team functions, and vendor/customer relationships. The evolvement of traditional base scope services to the integrated outage concept is addressed and includes discussions on changing customer needs, shared risks, and a partnership approach to outages. Experiences with concept implementation from a single service in 1984 to the current volume of integrated outage management presented in this paper. We at Westinghouse believe that the operators of nuclear power plants will continue to be aggressively challenged in the next decade to improve the operating and financial performance of their units. More and more customers in the U. S. are looking towards integrated outage as the way to meet these challenges of the 1990s, an arrangement that is best implemented through a long-term partnering with a single-source supplier of high quality nuclear and turbine generator outage services. This availability, and other important parameters

  1. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  2. Design of nuclear power plants

    International Nuclear Information System (INIS)

    The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.)

  3. Information technology for nuclear plant

    International Nuclear Information System (INIS)

    Sixteen papers are included. The first session covers the collection, storage and processing of relevant data (configuration management, the NEA reporting system and reliability data collection). The second group of papers is concerned with training and emergency control including simulation. The third session concerned plant control and maintenance including hardware and software problems. The final session is about artificial intelligence and expert systems applied to industrial management and control. An additional paper is about improved information exchange among nuclear operators. All papers are indexed separately. (UK)

  4. Nuclear power plant simulator software

    International Nuclear Information System (INIS)

    The development is described of software for the simulator of technological equipment of a WWER-440 nuclear power plant unit, using an SM 4-20 computer operating under DOS RV V2 and using a FORTRAN translator. An asynchronous interrupt system is used for kayboard control. All utilities for I/0 operations are written in the macroassembler. The basic concept of the software is shown schematically as are the relations between the program modules and the instructor and student environment. (J.B.). 1 fig

  5. Slovak Electric, plc, Mochovce Nuclear Power Plant

    International Nuclear Information System (INIS)

    In this popular scientific brochure a brief description of construction scheme of Bohunice Nuclear Power Plant is presented. Electricity generation in a nuclear power plant is described. Instrumentation and control system as well as nuclear safety principles applied on the NPP are presented

  6. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

  7. Statistical analysis about corrosion in nuclear power plants

    International Nuclear Information System (INIS)

    Nowadays, it has been carried out the investigations related with the structure degradation mechanisms, systems or and components in the nuclear power plants, since a lot of the involved processes are the responsible of the reliability of these ones, of the integrity of their components, of the safety aspects and others. This work presents the statistics of the studies related with materials corrosion in its wide variety and specific mechanisms. These exist at world level in the PWR, BWR, and WWER reactors, analysing the AIRS (Advanced Incident Reporting System) during the period between 1993-1998 in the two first plants in during the period between 1982-1995 for the WWER. The factors identification allows characterize them as those which apply, they are what have happen by the presence of some corrosion mechanism. Those which not apply, these are due to incidental by natural factors, mechanical failures and human errors. Finally, the total number of cases analysed, they correspond to the total cases which apply and not apply. (Author)

  8. Fukushima Nuclear Power Plant Accident and Nuclear Physicists

    Directory of Open Access Journals (Sweden)

    Otsuka Takaharu

    2014-03-01

    Full Text Available I give an overview on the Fukushima Nuclear Power Plant Accident and a report on voluntary activities of Japanese nuclear physicists in this terrible event, including their major outcome.

  9. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  10. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  11. Site preparation and excavation for foundation of power plant in construction works for No.2 plant in Tsuruga Nuclear Power Station

    International Nuclear Information System (INIS)

    Japan Atomic Power Co. has completed the improved Calderhall type power station in 1966, BWR type No.1 plant in Tsuruga Power Station in 1970, and the BWR type plant in Tokai No.2 Power Station in 1978. Now, as the fourth nuclear power plant, No.2 plant with 1160 MW output in Tsuruga Power Station is a PWR plant, and a prestressed concrete containment vessel was adopted for the first time in Japan. The site of No.2 plant is the limited area between No.1 plant and the Fugen Power Station, in which only the structures of the power plant can be placed, but there was no space for the temporary structure for construction works and abandoning earth. In excavation, the amount of special explosive was limited to avoid the effect of vibration and flying stones. The geological features of the site were examined by various tests. The outline of the civil engineering works is explained. The area of the site is 71,000 m2, and its level is EL + 7.0m. The amount of excavation is 770,000 m3. The excavation for the foundations of three main buildings amounts to 570,000 m3, and the amount of concrete is 17,500 m3. The method of excavation was selected according to the class of rocks. The excavation works were completed without trouble, and the construction is in progress smoothly. (Kako, I.)

  12. Multiprocessing in nuclear plant simulation

    International Nuclear Information System (INIS)

    This thesis describes the development of a multiprocessor for Nuclear Plant Simulation. A summary of the main features of continuous system simulation languages is presented. These languages simplify the development of simulation models by allowing the user to specify his program in a form closely related to its mathematical formulation. Existing computer architectures are studied for their suitability for the heavy computing requirements of simulation models. These are not well suited to the characteristics of simulation and hence do not yield the required performance. A novel computer architecture is described. This architecture is specially designed to match the characteristics of nuclear plant simulation. The author has developed a simulation language for the multiprocessor and has written a compiler for the language. The compiler has two special passes to analyse and partition the model so that the user can run programs on the multiprocessor transparently. Results from running real models on the multiprocessor have demonstrated the potential of the architecture and highlighted areas for future developments. (author)

  13. Comparative considerations on radioactive wastes, and on environmental aesthetical and thermal pollution caused in the normal operation of the main reactor types used in nuclear power plants

    International Nuclear Information System (INIS)

    The paper analyses four types of nuclear reactors, currently used at a world-wide scale from the point of view of the environmental impact due to the radioactive wastes and the thermal and aesthetical pollution. Solid, liquid, gaseous and tritide radioactive products are taken into consideration in case of normal operation conditions of the nuclear power plants. The comparison among the reactor types PWR, BWR, PHWR and HTGR shows a definite lagging for the case of the PHWR reactor, installed at the Cernavoda Nuclear Power Plant. (author). 7 figs., 1 tabs., 12 refs

  14. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  15. Nuclear Plant/Hydrogen Plant Safety: Issues and Approaches

    Energy Technology Data Exchange (ETDEWEB)

    Steven R. Sherman

    2007-06-01

    The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.

  16. Nuclear power plants and the environment

    International Nuclear Information System (INIS)

    These proceedings reproduce the papers presented at the 6th Colloquium of the French Environment Law Society and a summary of the ensuing discussions. The topics covered include a comparative analysis of nuclear and environment law, pollution law and nuclear waste, as well as presentations on the setting up and operation of nuclear power plants, nuclear safety, liability and insurance. (NEA)

  17. TOSHIBA CAE system for nuclear power plant

    International Nuclear Information System (INIS)

    TOSHIBA aims to secure safety, increase reliability and improve efficiency through the engineering for nuclear power plant using Computer Aided Engineering (CAE). TOSHIBA CAE system for nuclear power plant consists of numbers of sub-systems which had been integrated centering around the Nuclear Power Plant Engineering Data Base (PDBMS) and covers all stage of engineering for nuclear power plant from project management, design, manufacturing, construction to operating plant service and preventive maintenance as it were 'Plant Life-Cycle CAE System'. In recent years, TOSHIBA has been devoting to extend the system for integrated intelligent CAE system with state-of-the-art computer technologies such as computer graphics and artificial intelligence. This paper shows the outline of CAE system for nuclear power plant in TOSHIBA. (author)

  18. Nuclear Power Plant Lifetime Management Study (I)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

  19. IAEA provisional code of practice on management of radioactive waste from nuclear power plants

    International Nuclear Information System (INIS)

    This Code of Practice defines the minimum requirements for operations and design of structures, systems and components important for management of wastes from thermal nuclear power plants. It emphasizes what safety requirements shall be met rather than specifies how these requirements can be met; the latter aspect is covered in Safety Guides. The Code defines the need for a Government to assume responsibility for regulating waste management practices in conjunction with the regulation of a nuclear power plant. The Code does not prejudge the organization of the regulatory authority, which may differ from one Member State to another, and may involve more than one body. Similarly, the Code does not deal specifically with the functions of a regulatory authority responsible for such matters, although it may be of value to Member States in providing a basis for consideration of such functions. The Code deals with the entire management system for all wastes from nuclear power plants embodying thermal reactors including PWR, BWR, HWR and HTGR technologies. Topics included are: design, normal and abnormal operation, and regulation of management systems for gaseous, liquid and solid wastes, including decommissioning wastes. The Code includes measures to be taken with regard to the wastes arising from spent fuel management at nuclear power plants. However, the options for further management of spent fuel are only outlined since it is the subject of decisions by individual Member States. The Code does not require that an option(s) be decided upon prior to construction or operation of a nuclear power plant

  20. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Quarterly reports on the operation of Finnish nuclear power plants describe events and observations, relating to nuclear and radiation safety, which the Finnish Centre for Radiation and Nuclear Safety considers safety significant. Also other events of general interest are reported. The reports also include a summary of the radiation safety of plant personnel and the environment, as well as tabulated data on the plants' production and load factors

  1. Safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Experience shows that German nuclear power plants have always been operated reliably and safely. Over the years, the safety level in these plants has been raised considerably so that they can stand any comparison with other countries. This is confirmed by the two reports published by the Federal Ministry for the Environment on the nuclear safety convention. Behind this, there must obviously stand countless appropriate 'good practices' and a safety management system in nuclear power plants. (orig.)

  2. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Quarterly reports on the operation of Finnish nuclear power plants describe events and observations, relating to nuclear and radiation safety, which the Finnish Centre for Radiation and Nuclear Safety considers significant. Also other events of general interest are reported. The reports also include a summary of the radiation safety of plant personnel and the environment, as well as tabulated data on the plants' production and load factors

  3. Statistical analysis of fire events at US nuclear power plants

    International Nuclear Information System (INIS)

    The concern about fires as a potential agent of common cause failure in NPPs has greatly increased since the Browns Ferry NPP fire. Several regulatory actions were initiated following this incident. In investigating the chances of fire incident leading to core melt it is found that the unconditional frequency is about 1x10 incidents per reactor-year. The detailed reviews of fire events at nuclear plants are used in quantifying fire occurrence frequency required to carry out fire risk assessment. In this work the results of a statistical analysis of 354 fire incidents at US NPPs in the period from January 1965 to June 1985 are presented to quantify fire occurrence frequency. The distribution of fire incidents between the different types of NPPs (PWR, BWR or HTGR), the mode of plant operation, the probable cause of fire, the type of detectors detect the incident, who extinguished the fire, suppression equipment, suppression agent, the initiating combustible, the component or components affected by fire are all analysed for the studied 354 fire incidents. More than 50% of the incidents occurred during the construction phase, in many of them there is neither nuclear problem nor any safety problem, however these incidents delayed the startup of the units up to 2 years as happened in Indian Point unit 2 (1971). There are four major fire incidents at US NPPS in the first period of the study (1965-1978), not one of them in the last seven years (1979-1985) which clarify the development in the fire protection measures and technology. The fire events in US (NPPS) can be summarized in about 354 incidents at 33 locations due to 38 causes of fire with 0.17 fire events/plant/year

  4. Study on alternatives of inertisation of nuclear power plant containment

    International Nuclear Information System (INIS)

    In the course of a severe accident in a nuclear power plant, the hydrogen generation and other flammable gases, during the core degradation phase and the interaction corium-concrete, could produce the failure of the containment by overpressure of by combustion. According to the analysis of the potential effects of hydrogen evolution, following accidents inside the containment trough a Defense-in depth principle, which attempts to assure that the containment must not fail catastrophically, two techniques have been evaluated: a: Inertisation pre-accident and b: Inertisation post-accident. The technique of inertisation pre-accident consists in replacing the air of the containment with inert-gas like nitrogen (N2) or carbon dioxide (CO2) during the normal operation. The inertisation post-accident in combination with early venting system consists in replacing the air of the containment with inert-gas like nitrogen (N2) or carbon dioxide (CO2), immediately after the beginning of the accident, while the radioactivity is still negligible inside the containment. A system of inertisation pre-accident with nitrogen is used on BWR Mark I and Mark II. Investigations on the inertisation post-accident of the containment atmosphere during severe accidents have been carried out with different objectives from principles of the decade of 1980. Studies concerning hydrogen problem for the nuclear power plants Atucha I and CAREM-25 have permitted to know that the hydrogen generation during an accidental sequence with core degradation, would result important, being able to arrive to form explosive mixtures. In the present work, the applicability of the techniques of inertisation is analyzed for the containment of the Atucha I and CAREM-25, considering the particular design characteristics of these plants. (author)

  5. BWR fuel performance

    International Nuclear Information System (INIS)

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  6. Rock siting of nuclear power plants from a reactor safety standpoint

    International Nuclear Information System (INIS)

    The study has aimed at surveying the advantages and disadvantages of a rock sited nuclear power plant from a reactor safety standpoint. The studies performed are almost entirely concentrated on the BWR alternative. The design of a nuclear power plant in rock judged most appropriate has been studied in greater detail, and a relatively extensive safety analysis has been made. It is found that the presented technical design of the rock sited alternative is sufficiently advanced to form a basis for further projecting treatment. The chosen technical design of the reactor plant demands a cavern with a 45-50 metre span. Caverns without strengthening efforts with such spans are used in mines, but have no previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant. When locating a rock sited nuclear power plant, the same criteria must be considered as for an above ground plant, with additional stronger demands for rock quality. The presented rock sited nuclear power plant has been assessed to cost 20 % more in total construction costs than a corresponding above ground plant. The motivations for rock siting also depend on whether a condensing plant for only electricity production, or a plant for combined power production and district heating, is considered. The latter would under certain circumstances make rock siting look more attractive. (author)

  7. Electromagnetic compatibility of nuclear power plants

    International Nuclear Information System (INIS)

    Lately, there has been a mounting concern about the electromagnetic compatibility of nuclear-power-plant systems mainly because of the effects due to the nuclear electromagnetic pulse, and also because of the introduction of more-sophisticated and, therefore, more-susceptible solid-state devices into the plants. Questions have been raised about the adequacy of solid-state-device protection against plant electromagnetic-interference sources and transients due to the nuclear electromagnetic pulse. In this paper, the author briefly reviews the environment, and the coupling, susceptibility, and vulnerability assessment issues of commercial nuclear power plants

  8. Data retrieval techniques for nuclear power plants

    International Nuclear Information System (INIS)

    Data retrieval, processing retrieved data, and maintaining the plant documentation system to reflect the as-built condition of the plant are challenging tasks for most existing nuclear facilities. The information management systems available when these facilities were designed and constructed are archaic by today's standards. Today's plant documentation systems generally include hard copy drawings and text, drawings in various CAD formats, handwritten information, and incompatible databases. These existing plant documentation systems perpetuate inefficiency for the plant technical staff in the performance of their daily activities. This paper discusses data retrieval techniques and tools available to nuclear facilities to minimize the impacts of the existing plant documentation system on plant technical staff productivity

  9. Emergency evacuation plans for nuclear plants

    International Nuclear Information System (INIS)

    A critical commentary is presented on nuclear power plant emergency evacuation plans and procedures. The author alleges a deliberate withholding of information by officials in order to avoid generating public concern over power plant hazards

  10. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly at the study on the effects of the radiation in the materials of the reactor; a little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear tracks manufactured in the ININ is presented, for the environmental monitoring in penetrations around the primary container of the Unit 1 of the Laguna Verde power plant. The monitoring of neutrons carried out with ends of radiological protection, during those operational tests of the reactor. (author)

  11. Operating experience with nuclear power plants 2013

    International Nuclear Information System (INIS)

    The VGB Technical Committee 'Nuclear Plant Operation' has been exchanging operating experience about nuclear power plants for more than 30 years. Plant operators from several European countries are participating in the exchange. A report is given on the operating results achieved in 2013, events important to plant safety, special and relevant repair, and retrofit measures from Belgium, Germany, Finland, the Netherlands, Switzerland, and Spain. (orig.)

  12. Aging management guideline for commercial nuclear power plants-stationary batteries. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Berg, R.; Shao, J.; Krencicki, G.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

    1994-03-01

    The Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant stationary batteries important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  13. Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

    Energy Technology Data Exchange (ETDEWEB)

    Toman, G.; Gazdzinski, R.; O`Hearn, E. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

    1994-02-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  14. Aging Management Guideline for commercial nuclear power plants: Electrical switchgear. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Toman, G.; Gazdzinski, R.; Schuler, K. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

    1993-07-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of this AMG to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner which allows personnel responsible for performance analysis and maintenance, to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  15. Aging Management Guideline for commercial nuclear power plants: Battery chargers, inverters and uninterruptible power supplies

    International Nuclear Information System (INIS)

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant battery chargers, inverters and uninterruptible power supplies important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already, experienced) and aging management program activities to the more generic results and recommendations presented herein

  16. Aging management guideline for commercial nuclear power plants-stationary batteries

    International Nuclear Information System (INIS)

    The Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant stationary batteries important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  17. Aging Management Guideline for commercial nuclear power plants: Motor control centers

    International Nuclear Information System (INIS)

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  18. Nuclear power plant operating experience, 1976

    International Nuclear Information System (INIS)

    This report is the third in a series of reports issued annually that summarize the operating experience of U.S. nuclear power plants in commercial operation. Power generation statistics, plant outages, reportable occurrences, fuel element performance, occupational radiation exposure and radioactive effluents for each plant are presented. Summary highlights of these areas are discussed. The report includes 1976 data from 55 plants--23 boiling water reactor plants and 32 pressurized water reactor plants

  19. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Quarterly reports on the operation of Finnish nuclear power plants describe events and observations, relating to nuclear safety and radiation protection which the Finnish Centre for Radiation and Nuclear Safety considers safety significant. Safety-enhancing modifications at the nuclear power plants and issues relating to the use of nuclear energy which are of general interest are also reported. The reports include a summary of the radiation safety of plant personnel and the environment, as well as tabulated data on the production and load factors of the plants. In the first quarter of 1993, a primary feedwater system pipe break occurred at Loviisa 2, in a section of piping after a feedwater pump. The break was erosion-corrosion induced. Repairs and inspections interrupted power generation for seven days. On the International Nuclear Event Scale the event is classified as a level 2 incident. Other events in the first quarter of 1993 had no bearing on nuclear safety and radiation protection

  20. Dukovany nuclear power plant safety

    International Nuclear Information System (INIS)

    Presentation covers recommended safety issues for the Dukovany NPP which have been solved with satisfactory conclusions. Safety issues concerned include: radiation safety; nuclear safety; security; emergency preparedness; health protection at work; fire protection; environmental protection; chemical safety; technical safety. Quality assurance programs at all stages on NPP life time is described. Report includes description of NPP staff training provision, training simulator, emergency operating procedures, emergency preparedness, Year 2000 problem, inspections and life time management. Description of Dukovany Plant Safety Analysis Projects including integrity of the equipment, modernisation, equipment innovation and safety upgrading program show that this approach corresponds to the actual practice applied in EU countries, and fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of MORAWA Equipment Upgrading program

  1. Cofrentes nuclear power plant risk management tools

    International Nuclear Information System (INIS)

    Cofrentes Nuclear Power Plant (CNPP) is aware of tendencies and improvements in helping operator burden caused by accidents. CNPP recognizes the importance of being prepared to face any kind of accident and to improve the operation alike. For that reason CNPP is trying to get the available tools on the market and to adapt them to the plant as well as to build their own ones. These are reasons that drove CNPP to install the EPRI's Risk Monitor EOOS at the plant. The main features of this tool are: the use of plant specific PSA models, the re-quantification of the entire PSA in minutes, the quantitative and qualitative evaluations, the instantaneous and cumulative core damage frequency (CDF) and the evaluation of alternative maintenance configurations and strategies. The potential users of this tool are schedulers, operators and PSA analysts. About severe accident management, CNPP, as part of the Spanish-BWROG, is currently analyzing the generic products of the US-BWROG Accident Management Guidelines (AMG) in order to generate their specific ones. Also, in this BWR group, the development of tools to simulate accident scenarios beyond core damage will be studied and a training process oriented to warrant the optimum use of new EOPs/SAGs in accident scenarios will be implemented. As a starting point for a simulation tool development, CNPP has its specific model of some severe accident codes: to support IPE analysis a model for CNPP with MAAP3.0B was developed and qualified. The purpose of the model qualification was to check the correct operation of systems and equipment, together with the simulation reliability of the phenomena that occur in various scenarios. This was structured in four groups: operational transients; abnormal transients simulated previously with RETRAN; loss of coolant: accidents and sequences for checking the appropriate behavior of the drywell, suppression pool and vent models and containment systems. Other severe accident codes as CONTAIN and

  2. Program increasing of nuclear power plant safeness

    International Nuclear Information System (INIS)

    The results achieved within the project of national task 'Program increasing of nuclear power plant safeness' are presented in the document. The project was aimed to extend and deepen activities relating to safety increase of nuclear power units WWER-440 which play significant part in electricity production in the Slovak Republic. The application of advanced foreign calculating programs and calculation of radionuclide spreading in environment and techniques will influence the increase of extent, quality and international acceptance of safety analysis of nuclear power plant blocks WWER-440 and the risk valuation from operating nuclear power plants. Methodic resources for coping in emergency situation in nuclear energetics will be used for support in decision making in real time during radiation emergency on nuclear plant, region and state level. A long-term strategy in dealing with burnt fuel and radioactive substance formatting during nuclear power plant liquidation particularly with waste which is un acceptable in regional dump, has developed into a theoretical and practical preparation of solvable group for operating the converting centre Bohunice and in inactivating the nuclear power plant A-1. The diagnostic activities in nuclear power plants in the Slovak Republic have been elaborated into a project of norm documents in accordance with international norms for diagnostic systems. Presentation of new technologies and materials for repairs and reconstructions of components and nuclear power plant knots qualify increase in their reliability, safety and life. New objective methods and criterions for valuation and monitoring of the residual life and safety of fixed nuclear power plants. Results of problem solving linked with connecting the blocks of nuclear power plants to frequency regulation in electric network in the Slovak Republic are also presented in the document

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

    Science.gov (United States)

    Whitelaw, Robert L.

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

  4. Analysis of the noise of the jet pumps of the Unit 2 of the Laguna Verde nuclear power plant

    International Nuclear Information System (INIS)

    The use of the analysis of noise for the detection of badly functioning of the components of a BWR it is a powerful tool in the determination of abnormal conditions of operation, during the life of a nuclear plant of power. From the eighties, some nuclear reactors have presented problems related with the jet pumps and the knots of the recirculation. The Regulatory Commission of the United States, in the I E bulletin 80-07, recommended to carry out a periodic supervision of the pressure drop of the jet pumps, to prevent structural failures. In this work, methods of analysis of noise are used for the detection of abnormal conditions of operation of the jet pumps of a BWR. Signals are analysed to low and high frequency of pressure drop with the NOISE software that is in development. The obtained results show the behavior of the jet pumps of jet 6 and 11 before and after a partial blockade in their throats where the pump 6 return to their condition of previous operation and the pump 11 present a new fall of pressure, inside the limit them permissible of operation. The methodology of the analysis of noise demonstrated to be an useful tool for the badly functioning detection, and you could apply to create a database to supervise the dynamic behavior of the jet pumps of an BWR. (Author)

  5. Are atomic power plants saver than nuclear power plants

    International Nuclear Information System (INIS)

    It is rather impossible to establish nuclear power plants against the resistance of the population. To prevail over this resistance, a clarification of the citizens-initiatives motives which led to it will be necessary. This is to say: It is quite impossible for our population to understand what really heappens in nuclear power plants. They cannot identify themselves with nuclear power plants and thus feel very uncomfortable. As the total population feels the same way it is prepared for solidarity with the citizens-initiatives even if they believe in the necessity of nuclear power plants. Only an information-policy making transparent the social-psychological reasons of the population for being against nuclear power plants could be able to prevail over the resistance. More information about the technical procedures is not sufficient at all. (orig.)

  6. Condition monitoring of rotormachinery in nuclear power plants

    International Nuclear Information System (INIS)

    Due to safety and economical reasons diagnostic and monitoring systems are of growing interest in nuclear power plants and other complex industrial productions. Key components of NPP's are rotating machineries of the primary and secondary loops like PWR main coolant pumps, BWR recirculation pumps, turbines, fresh water pumps and feed water pumps. Diagnostic systems are requested which detect, diagnose and localize faulty operation conditions at an early stage in order to prevent severe failures and to enable predictive and condition oriented maintenance. The knowledge of characteristical machine signatures and their time dependent behaviour are the basis of efficient condition monitoring of rotating machines. The performance of reference measurements are of importance for fault detection during operation by trend settings. The comparison with thresholds given by norms and standards is only a small section of available possibilities. Therefore, for each machinery own thresholds should be determined using statistical time values, spectra comparison, cepstrum analysis and correlation analysis for source localization corresponding to certain machine operation conditions. (author). 14 refs, 15 figs

  7. Strategy of plant concept development in Hitachi

    International Nuclear Information System (INIS)

    Hitachi contributes to the society in various fields of nuclear power such as the Light Water Reactor field, the Fast Breeder Reactor, the fuel cycle, and the medical treatment. Since the beginning of a first commercial operation of a BWR in Japan, Hitachi has constructed 20 units of BWR. Hitachi continues its efforts in achieving high reliability and large-scale output, and in 1996, it completed Advanced Boiling Water Reactor (ABWR) in cooperation with various BWR utilities, General Electric Company, and Toshiba Corporation. Hitachi has enhanced the ABWR technology further based on the above enough experience. The latest technologies were reflected in Hamaoka unit 5 and Shika unit 2 as the latest ABWR plants. The further upgrade technologies would been reflected in Shimane-3, Ohma-1 and Higashidori-1 as ABWR plants under planning. Hitachi obtains the chance to construct the nuclear power plant continuously. For the next generation, Hitachi is working on developing nuclear power plants that take diversified needs and global characteristics into account. As one of the approaches, the output series formation of BWR is extended as the following. ABWR-II and ESBWR (Economic and Simplified BWR) as a large-scale centralized power supply emphasizing cost efficiency, Medium size ABWR and natural circulation type BWR as medium and small-scale distributed power supply that features flexibility to various market needs, such as minimized capital risks, timely return on a capital investment, etc. As another approach, Hitachi tries to extend the Light Water Reactor technology. RBWR (Resource-Renewable BWR ) that achieves a high conversion ratio over 1.0 based on the BWR technology will make the fuel cycle flexible. Hitachi will continue the challenge for the next ABWR and the future with the enough experience of BWR construction. (author)

  8. Medical consequences of a nuclear plant accident

    International Nuclear Information System (INIS)

    The report gives background information concerning radiation and the biological medical effects and damages caused by radiation. The report also discusses nuclear power plant accidents and efforts from the medical service in the case of a nuclear power plant accident. (L.F.)

  9. Quality assurance in nuclear power plant

    International Nuclear Information System (INIS)

    The factors related to the licensing procedures of a nuclear power plant (quality assurance and safety analysis) are presented and discussed. The consequences of inadequate attitudes towards these factors are shown and suggestions to assure the safety of nuclear power plants in Brazil are presented. (E.G.)

  10. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    In the third quarter of 1993, all of Finland's four nuclear power plant units were in power operation, with the exception of the annual maintenance outages of the Loviisa units. The load factor average of the plant units was 83.6 %. None of the events which occurred during this annual quarter had any bearing on nuclear or radiation safety. (4 figs., 5 tabs.)

  11. Nuclear power plants for protecting the atmosphere

    International Nuclear Information System (INIS)

    Some figures are presented comparing date on the CO2 emission and oxygen consumption of nuclear, natural gas fired, advanced coal fired and oil fired power plants, for the same amounts of electricity generated. The data were deduced from the Paks Nuclear Power Plant, Hungary. (R.P.)

  12. Radiological characterization of nuclear plants under decommissioning

    International Nuclear Information System (INIS)

    In the present work a description of major problems encountered in qualitative and quantitative radiological characterization of nuclear plants for decommissioning and decontamination purpose is presented. Referring to several nuclear plant classes activation and contamination processes, direct and indirect radiological analysis and some italian significant experience are descripted

  13. Technical and economic proposal for the extension of the Laguna Verde Nuclear Power plant with an additional nuclear reactor

    International Nuclear Information System (INIS)

    The increment of the human activities in the industrial environments and of generation of electric power, through it burns it of fossil fuels, has brought as consequence an increase in the atmospheric concentrations of the calls greenhouse effect gases and, these in turn, serious repercussions about the environment and the quality of the alive beings life. The recent concern for the environment has provoked that industrialized countries and not industrialized carry out international agreements to mitigate the emission from these gases to the atmosphere. Our country, like part of the international community, not is exempt of this problem for what is necessary that programs begin guided toward the preservation of the environment. As for the electric power generation, it is indispensable to diversify the sources of primary energy; first, to knock down the dependence of the hydrocarbons and, second, to reduce the emission of polluting gases to the atmosphere. In this item, the nucleo electric energy not only has proven to be safe and competitive technical and economically, able to generate big quantities of electric power with a high plant factor and a considerable cost, but rather also, it is one of the energy sources that less pollutants it emits to the atmosphere. The main object of this work is to carry out a technical and economic proposal of the extension of the Laguna Verde Nuclear power plant (CNLV) with a new nuclear reactor of type A BWR (Advanced Boiling Water Reactor), evolutionary design of the BWR technology to which belong the two reactors installed at the moment in the plant, with the purpose of increasing the installed capacity of generation of the CNLV and of the Federal Commission of Electricity (CFE) with foundation in the sustainable development and guaranteeing the protection of the environment by means of the exploitation of a clean and sure technology that counts at the moment with around 12,000 year-reactor of operational experience in more of

  14. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  15. Spent fuel transport and storage system for NOK nuclear power plants

    International Nuclear Information System (INIS)

    Full text: Both NOK nuclear power plants in Switzerland - LEIBSTADT (KKL) BWR nuclear power plant and BEZNAU ( KKB) PWR nuclear power plant - have opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TNTM52L and TNTM97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNTM24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to KKL requirements to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TNTM52L and TNTM97L casks by the KKL and ZWILAG operators. In 2004, NOK also ordered three TNTM24 GB transport and storage casks for PWR fuel types. These casks are presently under manufacturing. TNTM52L: One TNTM52L unit was delivered to KKL in 2000. This TNTM52L cask has been used as a transport cask for the shipment of KKL spent fuel to COGEMA La Hague plant. After that, it has been used for the dry storage of KKL spent fuel at ZWILAG. The paper will present the wide range of operations made with the TNTM52L cask: loading, routine transport, unloading, maintenance, transport to ZWILAG and storage at ZWILAG. The experience gained from these operations will be presented. TNTM97L: Nine TNTM97L units were ordered by KKL. They have all been loaded and successfully stored at ZWILAG. The TNTM97L transport and storage cask has the highest payload ever designed world wide for BWR spent fuel. The paper will detail the operational feedback with this 'giant', from the cask loading at KKL to its transport and storage at ZWILAG. TNTM24 BHL: KKL has recently opted for the TNTM24BH cask design. The TNTM24BH was first designed for the fuel elements of the Swiss nuclear power plant of Muehleberg. By taking advantage of TNTM52L and TNTM97L experience, the TNTM24BHL (L for Leibstadt) will fulfil KKL future needs

  16. Transportation activities for BWR fuels at NFI

    International Nuclear Information System (INIS)

    Nuclear Fuel Industries, LTD. (NFI) supplies fuel assemblies for both PWR and BWR in Japan. We can also manage transportation of the fuel assemblies from our fabrication facilities to the nuclear power plants of Japanese utilities. For the transportation of fuel assembly, we designed and fabricated the transportation containers to meet the requirements of the IAEA regulations, and licensed in Japan. This paper introduced the recent activity and R and D of NFI concerning transportation of BWR fuel assembly. NT-XII transportation container was developed for fresh BWR fuel assemblies. NT-XII container consists of inner container and outer container. Two BWR fuel assemblies with up to 5 wt.-% 235-U enriched are enclosed in an inner container. In the concept of NT-XII container design, we made the best priority to transportation efficiency, as well as ensuring fuel integrity during transportation. NT-XII has been used since 2002 in Japan. Thanks to the lightening weight of containers, the number of containers to be loaded to one transportation truck was increased up to 9 containers (equivalent to 18 fuel assemblies) compared with former type container (NT-IV transportation container) which can be loaded up to 6 containers (equivalent to 12 fuel assemblies). In addition to the design of brand-new container, we promote the improvement of the packaging methods. In Japan, in order to reduce the damage to the fuel rod and fuel spacer while transporting, polyethylene sleeves which are called 'packing separators' are inserted in the rod-to-rod gap of fuel assembly. However, packing separators requires time and cost for the installation at fuel fabrication facility. In the same way, huge time and cost are needed for removal of packing separators at nuclear power plant. For the improvement of preparation efficiency before and after transportation, we investigated the influence of vibration to fuel integrity in case of transportation without packing separators. Based on the above

  17. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    The Finnish nuclear power plant units Loviisa 1 and 2 as well as TVO I and II were in operation for almost all the time in the first quarter of 1992. The load factor average was 99.8%. All events which are classified on the International Nuclear Event Scale were level 0/below scale on the Scale. Occupational radiation doses and releases of radioactive material off-site remained well below authorised limits. Only quantities of radioactive material insignificant to radiation exposure, originating from the nuclear power plants, were detected in samples collected in the vicinity of the nuclear power plants

  18. Nuclear power plants for mobile applications

    Science.gov (United States)

    Anderson, J. L.

    1972-01-01

    Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. The technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants are examined. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

  19. Plant nuclear proteomics for unraveling physiological function.

    Science.gov (United States)

    Yin, Xiaojian; Komatsu, Setsuko

    2016-09-25

    The nucleus is the subcellular organelle that functions as the regulatory hub of the cell and is responsible for regulating several critical cellular functions, including cell proliferation, gene expression, and cell survival. Nuclear proteomics is a useful approach for investigating the mechanisms underlying plant responses to abiotic stresses, including protein-protein interactions, enzyme activities, and post-translational modifications. Among abiotic stresses, flooding is a major limiting factor for plant growth and yields, particularly for soybean. In this review, plant nuclei purification methods, modifications of plant nuclear proteins, and recent contributions to the field of plant nuclear proteomics are summarized. In addition, to reveal the upstream regulating mechanisms controlling soybean responses to flooding stress, the functions of flooding-responsive nuclear proteins are reviewed based on the results of nuclear proteomic analysis of soybean in the early stages of flooding stress. PMID:27004615

  20. Occupational exposure management at nuclear power plants

    International Nuclear Information System (INIS)

    The Information System on Occupational Exposure (ISOE), a joint initiative of the OECD Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA), has become a unique worldwide programme on the protection of workers in nuclear power plants, including a network for the exchange of experience in the area of occupational exposure management, and the world's largest database on occupational exposure from nuclear power plants. Each year, an international workshop or symposium offers a forum for radiation protection professionals from the nuclear industry, operating organisations and regulatory authorities to exchange information on practical experience with occupational radiation exposure issues in nuclear power plants. These proceedings include the presentations made at the Third ISOE European Workshop on Occupational Exposure Management at Nuclear Power Plants, held in April 2002 in Portoroz, Slovenia. (author)

  1. Super-simulator for nuclear power plant

    International Nuclear Information System (INIS)

    The super-simulator is a conceptual name for future innovative simulators of nuclear power plants which surpass, to a large extent, the capabilities and performances of existing nuclear power plant simulators for operator training, plant analyzers for engineering studies or computer codes for dynamics analysis. Such super-simulators will be realized by employing highly advanced methods of mathematical modeling of the physical chemical and other related processes in the nuclear power plants and innovative methods of numerical computation using rapidly evolving high performance computing systems. (author)

  2. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  3. Nuclear

    International Nuclear Information System (INIS)

    This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)

  4. Fire protection at nuclear power plants

    International Nuclear Information System (INIS)

    The guide presents specific requirements for the design and implementation of fire protection arrangements at nuclear power plants and for the documents relating to the fire protection that are to be submitted to STUK (Finnish Radiation and Nuclear Safety Authority). Inspections of the fire protection arrangements to be conducted by STUK during the construction and operation of the power plants are also described in this guide. The guide can also be followed at other nuclear facilities

  5. Academic training for nuclear power plant operators

    International Nuclear Information System (INIS)

    After having briefly outlined the importance of academic training of nuclear power plant operators and the objectives of such a training (quality, addressing industry needs), the author evokes the programs implemented at the Center for Nuclear Studies of the Memphis State university. He notices that an academic degree is necessary for the recognition of the operator's job as a professional, and that such a training program is useful to improve safety and reliability of nuclear power plants

  6. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    In the Quarterly Reports on the operation of the Finnish nuclear power plants such events and observations are described relating to nuclear and radiation safety which the Finnish Centre for Radiation and Nuclear Safety considers safety significant. Also other events of general interest are reported. The report also includes a summary of the radiation safety of the plants' workers and the environment, as well as tabulated data on the production and load factors of the plants. The Finnish nuclear power plant units Loviisa 1 and 2 as well as TVO I and II were in commercial operation during the whole first quarter of 1991. The load factor average was 99.1 %. Failures have been detected in the uppermost spacing lattices of nuclear fuel bundles removed from the Loviisa nuclear reactors. Further investigations into the significance of the failures have been initiated. In this quarter, renewed cooling systems for the instrumentation area were introduced at Loviisa 1. The modifications made in the systems serve to ensure reliable cooling of the area even during the hottest summer months when the possibility exists that the temperature of the automation equipment could rise too high causing malfunctions which could endanger plant safety. Occupational radiation doses and external releases of radioactivity were below prescribed limits in this quarter. Only small amounts of radioactive substances originating in nuclear power plants were detected in samples taken in the vicinity of nuclear power plants

  7. Elecnuc. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1999 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1999; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; grid connection forecasts; world electric power market; electronuclear owners and share holders in EU, capacity and load factor; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; 1999 gross load factor by operator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

  8. Operating experience from Swedish nuclear power plants 2007

    International Nuclear Information System (INIS)

    The overall safety level of the Swedish nuclear power plants during 2007 has been high, not only in a national, but also in an international, perspective. There were no serious disturbances that could pose a threat to safety. However, in terms of electricity production, 2007 was not among the best years: at 64.3 TWh, electricity production was the lowest for five years. The total average energy availability was lower than in previous years. For the BWR reactors, it amounted to 81 %, as against an average value of about 83 % for the last fi ve years. For the PWRs, the difference was greater, with an energy availability of 80.8 % in 2007, considerably less than the fi ve-year average of 88.7 %. Rather too many, and long, refuelling outages due to extensive modernisation and conversion works, including the turbine at Ringhals 3, in combination with the knock-on effects of the Forsmark incident in 2006, contributed to the overall declining availability. Other factors included fuel damage at Oskarshamn 3 and replacement of rubber membrane seals at Forsmark 1 and 2. During the year, KSU started up a new training facility for the maintenance staff of all Swedish power plants at the shut-down Barsebaeck site. It enables electrical and mechanical maintenance personnel to be trained in authentic conditions, working on real components and systems before doing the same under live conditions at their home power plants. The autumn of 2007 was devoted to development of training objectives and programmes, with actual training starting in the spring of 2008

  9. Human factors in nuclear power plant operations

    International Nuclear Information System (INIS)

    This report describes some of the human factors problems in nuclear power plants and the technology that can be employed to reduce those problems. Many of the changes to improve the human factors in existing plants are inexpensive, and the expected gain in human reliability is substantial. The human factors technology is well-established and there are practitioners in most countries that have nuclear power plants

  10. Human factors in nuclear power plants

    International Nuclear Information System (INIS)

    This report describes some of the human factors problems in nuclear power plants and the technology that can be employed to reduce those problems. Many of the changes to improve the human factors in existing plants are inexpensive, and the expected gain in human reliability is substantial. The human factors technology is well-established and there are practitioners in most countries that have nuclear power plants. (orig.)

  11. Reducing radiation levels at boiling water reactors of a commercial nuclear power plant fleet

    International Nuclear Information System (INIS)

    Boiling Water Reactors (BWRs) have suffered from high radiation fields in the primary loop, typically measured by the 'BRAC' (BWR Radiation Level Assessment and Control) reactor recirculation system (RRS) dose rates. Reactor water chemistry and activated corrosion product measurements are important in understanding changes in radiation fields in components and systems of a BWR. Several studies have been conducted at Exelon Nuclear's 14 BWRs in order to understand more fully the cause and effect relationships between reactor water radioactive species and radiation levels. Various radiation control strategies are utilized to control and reduce radiation levels. The proper measurement of radioactive soluble and insoluble species is a critical component in understanding radiation fields. Other factors that impact radiation fields include: noble metal applications; hydrogen injection; zinc addition; chemistry results; cobalt source term; fuel design and operation. Chemistry and radiation field trending and projections are important tools that assist in assessing the potential for increased radiation fields and aiding outage planning efforts, including techniques to minimize outage dose. This paper will present the findings from various studies and predictor tools as well as provide recommendations for continued research efforts in this field. Current plant data will be shared on reactor water radioactive species, plant radiation levels, zinc addition amounts and other chemistry controls. (author)

  12. Nuclear Power Plants in the World

    International Nuclear Information System (INIS)

    The Japan Atomic Industrial Forum (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Developmental trends on nuclear power plants in the world'. In this report, some data at the end of 1999 was made up on bases of answers on questionnaires from 72 electric companies in 31 nations and regions in the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; location of the plants; the plants in the world; and so forth. And, it also has some survey results on the 'Liberalization of electric power markets and nuclear power generation' such as some 70% of respondents in nuclear power for future option, gas-thermal power seen as power source with most to gain from liberalization, merits on nuclear power generation (environmental considerations and supply stability), most commonly voiced concern about new plant orders in poor economy, and so forth. (G.K.)

  13. Nuclear plant data systems - some emerging directions

    International Nuclear Information System (INIS)

    Significant changes have occurred in recent years in the nuclear power industry to accentuate the need for data systems to support information flow and decision making. Economic conditions resulting in rapid inflation and larger investments in new and existing plants and the need to plan further ahead are primary factors. Government policies concerning environmental control, as well as minimizing risk to the public through increased nuclear safety regulations on operating plants are additional factors. The impact of computer technology on plant data systems, evolution of corporate and plant infrastructures, future plant data, system designs and benefits, and decision making capabilities and data usage support are discussed. (U.K.)

  14. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

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

  15. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B4C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  16. Pathways and estimated consequences of radionuclide releases from a nuclear power plant

    International Nuclear Information System (INIS)

    The radioactive corrosion products 60Co, 58Co, 54Mn, 65Zn, 51Cr and 110Agsp(m) which are released from Barsebaeck Nuclear Power Plant (Sweden) into the marine environment have been studied. The brown seaweed Fucus vesiculosus has been found to be excellent bioindicator for these radionuclides. The distribution of activation products along the west coast of Sweden has been studied and can very well be described with a power function. Uptake and retention of the activation products in Fucus have been studied and results are reported. The activity concentration in Fucus well reflects the disharge rate from the power plant. Other bioindicators have been studied ; mainly the crustaceans Idothea and Gammarus. A simulation of 131I-release following a fictious BWR 1 accident is described as well as a model for estimation of the individual and collective dose equivalents arising from intake of contaminated milk. (Author)

  17. Maintenance of nuclear power plant

    International Nuclear Information System (INIS)

    Maintenance action of nuclear power plant (NPP) was described. Maintenance of NPP aimed at assurance of required function of NPP's equipment so as to prevent release of radioactive materials into the environment as well as attainment of stable operation of NPP. Philosophy of NPP safety was based on defense-in-depth or multiple barriers requiring specified function for the equipment. Preventive maintenance was essential to NPP's equipment and the scope of maintenance was decided on priority with adequate method and frequency of inspection. Most inspection was conducted during periodic inspection at outage. Repair or improvement works were performed if needed. Periodic inspection period was very long and then capacity factor of NPP was low in Japan compared with foreign data although frequency of unscheduled shutdown was very low. Introduction of reability- centered maintenance was requested based on past experiences of overhaul inspection. Technical evaluation of aged NPP had been conducted on aging phenomena and promotion of advanced maintenance was more needed. (T. Tanaka)

  18. Los Alamos Nuclear plant analyzer

    International Nuclear Information System (INIS)

    The Relational Database software obtained from Idaho National Engineering Laboratory is implemented on the Los Alamos Cray computer system. For the Nuclear Plant Analyzer (NPA), Los Alamos retained a graphics display terminal and a separate forms terminal for mutual compatibility, but integrated both the terminals into a single-line full-duplex mode of communications, using a single keyboard for input. Work on the program-selection phase of an NPA session is well underway. The final phase of implementation will be the Worker or graphics-driver phase. The Los Alamos in-house NPA has been in use for some time, and has given good results in analyses of four transients. The NPA hydrocode has been developed in to a fast-running code. The authors have observed an average of a factor-of-3 speed increase for typical slow reactor-safety transients when employing the stability enhancing two-step (SETS) method in the one-dimensional components using PF1/MOD1. The SETS method allows violation of the material Courant time-step stability limit and is thus stable at large time steps. The SETS method to the three-dimensional VESSEL component in the NPA hydrocode has been adapted. In addition to the speed increase from this new software, significant additional speed is expected as a result of new hardware that provides for vectorization or parallelization

  19. Citizens contra nuclear power plants

    International Nuclear Information System (INIS)

    Is Wyhl the beginning of a new citizens' movement against official policies concerning atomic energy or is it the end of citizens' initiatives of latter years. Did democracy pass its test in Wyhl, or was the state's authority undermined. The danger of atomic energy was not the only concern of the citizens of the Rhine valley who demonstrated against the planned nuclear power plant, but also the quality of industrial and energy planning in which the democratic foundations have to be safeguarded. In the meantime, the doubts increase that this source of energy is of a not dangerous nature, and the myth of supposedly cheap atomic energy has been scattered. The dangers in connection with waste transport and storage were made public beyond the boundaries of the places in question, in particular as a result of the demonstrations. The publication documents the course of the demonstration and the site occupation from the beginning of Febuary 1975 onwards. The occupation still continued when the booklet was published despite the decision of the Administrative Court in Freiburg at the end of March (prohibition of commencement of building until the verdict on the principal suit against the overall project has been reached, the final decision to be made by the Higher Administrative Court in Mannheim). The author aims at describing the new quality of citizens' commitments in this booklet. (orig./LN)

  20. Capturing stranded value in nuclear plant assets

    International Nuclear Information System (INIS)

    The New York Power Authority owns and operates two nuclear plants - generating assets that demand innovative approaches in an increasingly competitive market for electricity. NYPA plans to turn to the specialized skills of one of the industry's best nuclear operators, with the goal of maximizing the plants' value and yielding lower cost electricity for its customers. In the transition to a competitive electric power business, much of the power industry's focus has been directed at the issue of stranded assets, particularly those associated with investments in nuclear plants. Many nuclear plants are struggling with the twin disabilities of relatively low and/or variable operating performance and high costs. The result is a cost for power that may not be fully recovered in the marketplace, thus stranding part or all of the asset's value. At the NYPA attention has been on the flip side of stranded costs: what they term the stranded value associated with nuclear plants. They view stranded value as the potential additional value - increased power generation and reduced costs, while maintaining high levels of safe operations - that could be obtained from an average nuclear plant if it could be operated at the performance levels attained by the best neclear plant operators in the country. They do not accept current performance or industry averages as a given. They expect to deliver the full potential value of their nuclear assets to their customers by implementing a new operating strategy

  1. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    Science.gov (United States)

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.

    2012-05-01

    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

  2. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    In 1985 an incident at Toledo Edison's Davis Besse plant caused the U.S. Nuclear Regulatory Commission (NRC) to re-evaluate the technical information that the utilities had readily available to support the design of their plants. The Design Basis programs, currently on going in most U.S. utilities, have been the nuclear industry's response to the needs identified by this re-evaluation. In order to understand the Design Basis programs which have been implemented by the U.S. nuclear utilities, it is necessary to understand the problem as it was perceived by the nuclear industry (the utilities, the original NSSS designers and the regulators) after the Davis-Besse incident, the subsequent programs undertaken by the industry under the leadership of INPO and NUMARC, the NRC's actions, and the overall evolution of the industry's vision in relation to this problem. This paper presents the history of the design basis efforts from the first recognition of the problem by the NRC after the Davis-Besse incident, describes the actions taken by the NRC, INPO, NUMARC, the U.S. utilities and the NSSS designers, and brings the problem statement up-to-date in relation to the vision presently held by the U.S. nuclear industry. It then presents a technical discussion to develop a detailed definition of design basis information to support the problem statement. The information originally supplied by the NSSS designers during the plant design and construction is discussed as well as its relationship to the previously defined design basis information. This section of the paper concludes by defining the additional information needed by nuclear utilities to satisfy the requirements developed from the problem statement. Having developed a definition of the additional information (i.e., information not originally supplied during design and construction) required to solve the design basis problem as it is presently perceived by the U.S. nuclear industry, the paper then discusses design basis

  3. Development of BWR ex-core nuclear instrumentation system. Experimental verification of ex-core local power monitoring method by using neutron streaming pipes

    International Nuclear Information System (INIS)

    We study an ex-core nuclear instrumentation system for a next generation Boiling Water Reactor (BWR). For an example of existing BWRs, in an Advanced BWR (ABWR), about 210 in-core sensors are installed. About sixty penetrations of Reactor Pressure Vessel (RPV) are built for these signal cables. The core monitoring with ex-core sensors instead of in-core sensors reduces the RPV penetrations and extends the sensor life. To achieve this, we created a local power monitoring method with Neutron Streaming Pipes (NSPs), where local power is directly determined from ex-core neutrons through NSPs inserted into the core. In the previous work, applicability of this method for the ABWR was analytically shown. The most critical parameter in this evaluation was the ex-core neutron flux through the NSP. Hence, we experimentally verified validity of the previous analytical values of the flux in a research reactor, Toshiba Nuclear Critical Assembly. The verification was performed in two respects. Firstly, fundamental characteristics of the ex-core neutron flux through a NSP that do not depend on the reactor types were confirmed by measuring ex-core neutron count rates. (1) The ex-core neutron flux φ does not depend on the NSP wall thickness, when the thickness is on the order of millimeters (the measured values: independent within experimental error of 3.0% at 1 σ). (2) The ex-core neutron streaming rate (= φ x the inner cross-sectional area of the NSP) is proportional to the fourth power of the NSP inner diameter (the measured values: the 3.7 and 3.8 power). Secondly, reliability of the neutron streaming calculation was verified by comparing two independent calculations of a number of the incident ex-core neutrons at the detector. One was a neutron streaming calculation with in-core gold foil activation method. The other was a detector response calculation with measured ex-core neutron count rates. They were in good agreement within experimental and analytical error of 7.8% at

  4. Qualification, certification and training systems of a Japanese nuclear power plant supplier

    International Nuclear Information System (INIS)

    Training and indoctrination of personnel are very important for performing quality assurance activities in nuclear power plants. The paper presents policies, procedures and practices with regard to a training system for site construction workers, a training system for plant operation personnel at a special facility, and services and activities for operating plants. The training system for site workers and technicians offers basic training for all workers, and special training and qualification for field welders, cable-termination workers and non-destructive examination personnel. In addition to the qualification system for field welders applied by the regulatory body, a privat-company qualification system exists. Also, a training centre for BWR operators has been established. This facility has a simulator duplicating the main control room of an actual plant and a computer-assisted instruction system. Standard training courses, short-term basic courses, re-training courses, group training courses and special training courses are held at the training centre. Finally, the services and activities performed by Toshiba Corporation for operating plants are described. These activities of the plant supplier aim at keeping up and further enhancing the safety and reliability of operating plants and mainly consist of: (1) collection and evaluation of plant operation data, with re-evaluation and improvement of systems and components; (2) development of new maintenance techniques; (3) development of measures for reducing annual outage periods; and (4) plant emergency preparedness. (author)

  5. Methods of assessing nuclear power plant risks

    International Nuclear Information System (INIS)

    The concept of safety evalution is based on safety criteria -standards or set qualitative values of parameters and indices used in designing nuclear power plants, incorporating demands on the quality of equipment and operation of the plant, its siting and technical means for achieving nuclear safety. The concepts are presented of basic and optimal risk values. Factors are summed up indispensable for the evaluation of the nuclear power plant risk and the present world trend of evaluation based on probability is discussed. (J.C.)

  6. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    This general review of the operation of the Finnish nuclear power plants concentrates on such events and discoveries related to nuclear and radiation safety as the regulatory body, the Finnish Centre for Radiation and Nuclear Safety, regards as noteworthy. The report also includes a summary of the radiation safety of the personnel and the environment, as well as tubulated data on the production and load factors of the plants. In the report period, no event essentially degraded plant safety nor posed a radiation hazard to the personnel or the environment

  7. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    This general review of the operation of the Finnish nuclear power plants concentrates on such events and discoveries related to nuclear and radiation safety as the regulatory body, the Finnish Centre for Radiation and Nuclear Safety, regards as noteworthy. The report also includes a summary of the radiation safety of the personnel and the environment, as well as tabulated data on the production and load factors of the plants. In the report period, no event essentially degraded plant safety nor posed a radiation hzard to the personnel or the environment

  8. Nuclear power plants in Europe 1989

    International Nuclear Information System (INIS)

    The report covers the situation of the nuclear power plant industry in 18 European countries, seven belonging to the COMECON. There are at present 229 nuclear power plant units in operation, producing 168 087 MWe. Sixty-six units are under construction and will add a capacity of 62 355 MWe, and further planned 33 units are designed to supply 32 872 MWe. This results in a total of 328 nuclear power plant units with a capacity of 263 318 MWe. The number of units in operation in West European countries including Yugoslavia was 153 in April 1989, totalling a capacity of 122 809 MWe. (orig./UA)

  9. Safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    The role and purpose of safety principles for nuclear power plants are discussed. A brief information is presented on safety objectives as given in the INSAG documents. The possible linkage is discussed between the two mentioned elements of nuclear safety and safety culture. Safety culture is a rather new concept and there is more than one interpretation of the definition given by INSAG. The defence in depth is defined by INSAG as a fundamental principle of safety technology of nuclear power. Discussed is the overall strategy for safety measures, and features of nuclear power plants provided by the defence-in-depth concept. (Z.S.) 7 refs

  10. U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 7. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been recently expanded for BWR out-of-phase behavior. Out-of-phase oscillation is a phenomenon that occurs at BWRs. During this kind of event, half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. The HRS will be used for development and validation of stability monitoring and control techniques as part of an ongoing U.S. Department of Energy Nuclear Engineering Education and Research grant. The Penn State TRIGA reactor is used to simulate BWR fundamental mode power dynamics. The first harmonic mode power, together with detailed thermal hydraulics of boiling channels of both fundamental mode and first harmonic mode, is simulated digitally in real time with a computer. Simulations of boiling channels provide reactivity feedback to the TRIGA reactor, and the TRIGA reactor's power response is in turn fed into the channel simulations and the first harmonic mode power simulation. The combination of reactor power response and the simulated first harmonic power response with spatial distribution functions thus mimics the stability phenomena actually encountered in BWRs. The digital simulations of the boiling channels are performed by solving conservation equations for different regions in the channel with C-MEX S-functions. A fast three-dimensional (3-D) reactor power display of modal BWR power distribution was implemented using MATLAB graphics capability. Fundamental mode, first harmonic, together with the total power distribution over the reactor cross section, are displayed. Because of the large amount of computation for BWR boiling channel simulation and real-time data processing and graph generation, one computer is not sufficient to handle these jobs in the hybrid reactor simulation environment. A new three-computer setup has been

  11. Economic performance of nuclear plants: How competitive

    International Nuclear Information System (INIS)

    There is no uniquely correct cost for nuclear or other modes of electricity production, and there is little to be gained by seeking complete standardization of reference values used in making cost comparisons, for example, between nuclear and fossil-fuelled power plants. Even at the national level, such inter-fuel comparisons (nuclear versus fossil) have only limited ''generic'' value, due to the number of assumptions and operating conditions that are behind each example. Nevertheless, results from such studies can contribute to an improved understanding of the worldwide economic viability of nuclear power. Therefore, this article reviews some of the reported experience with nuclear power economic performance and estimates of future nuclear power costs, in comparison with fossil-fuel-fired plants. It is reemphasized, however, that the cost data presented should not be used as reference data for planning purposes, but are valid only to give an overall indication of the general economic competitiveness of nuclear power

  12. Study of multi cycles with FCS-II code for Unit 1 of Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    A study of 20 operating cycles for the BWR type reactor of Laguna Verde Nuclear Power Plant UNIT 1 is presented. The study was performed by the optimization group of fuel reloading by means of FCS-II code which is part of a computing package of Fuel Management System; with this results, part of the information concerning to a multi cycles analysis is evaluated; the information was provided by General Electric (GE) to Comision Federal de Electricidad. This study is part of the inter-institutional project of Fuel Management inside core for Laguna Verde Nuclear Power Plant where the involved institutions are Instituto de Investigaciones Electricas and Instituto Nacional de Investigaciones Nucleares under the direction of Comision Federal de Electricidad. (Author)

  13. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    During the second quarter of 1990 the Finnish nuclear plant units Loviisa 1 and 2 and TVO and II were in commercial operation for most of the time. The feedwater pipe rupture at Loviisa 1 and the resulting inspections and repairs at both Loviisa plant units brought about an outage the overall duration of which was 32 days. The annual maintenance outages of the TVO plant units were arranged during the report period and their combined duration was 31.5 days. Nuclear electricity accounted for 35.3% of the total Finnish electricity production during this quarter. The load factor average of the nuclear power plant units was 83.0%. Three events occurred during the report period which are classified as Level 1 on the International Nuclear Event Scale: feedwater pipe rupture at Loviisa 1, control rod withdrawal at TVO I in a test during an outage when the hydraulic scram system was rendered inoperable and erroneous fuel bundle transfers during control rod drives maintenance at TVO II. Other events during this quarter are classified as Level Zero (Below Scale) on the International Nuclear Event Scale. Occupational radiation doses and external releases of radioactivity were considerably below authorised limits. Only small amounts of nuclides originating in nuclear power plants were detected in samples taken in the vicinity of nuclear power plants

  14. A multivariate correlation method as a tool for fault detection and its application to feedwater system in a nuclear power plant

    International Nuclear Information System (INIS)

    A new method based on multivariate correlation has been developed for surveillance and anomaly detection in nuclear power plants. Natural signal noises observed under normal plant operations are applied in the method, thus the method is suitable for online surveillance. In the method, correlation of signals and noise properties in the process are estimated. Eigenvalues of normalized covariance matrix and correlation function among signals are used for surveillance of process dynamics and operational condition. The effectiveness of the method has been demonstrated with an application to feedwater system of a BWR plant. (author)

  15. Nuclear power plants 1985 - a world survey

    International Nuclear Information System (INIS)

    The Quick Statistics annually compiled by atw lists 355 nuclear power plants with 263,027 gross MWe in operation in 26 countries in late 1985. Another 163 units with 157801 MWe were under construction in 25 countries, and 75 units with 77,328 MWe were ordered in 18 countries. This adds up to a total of 593 units with an aggregate 498,156 MWe. In the course of 1985 30 nuclear generating units with 28,375 MWe were newly commissioned in 17 countries and 33 units in 14 countries started commercial operation. In the survey of electricity generation in 1985 for which no information was made available from the Eastern block countries, a total of 289 nuclear power plants were covered. In the year under review they generated an aggregate 1,270,854 GWh, which is 18.4% less than in the previous year. The highest nuclear generation again was recorded in the USA with 404,515 GWh, followed by France with 24,644 GWh. The KWG Grohnde Nuclear Power Station in the Federal Republic of Germany attained the maximum annual production figure of 1,477 GWh. The survey includes 10 tables indicating the generating performance of each nuclear power plant, the development of electricity generation in nuclear plants, and status of, nuclear power plants at the end of 1985 arranged by countries, types of reactors, and reactor manufacturers. (orig.)

  16. Nuclear power plants and the environment

    International Nuclear Information System (INIS)

    The environmental impacts are compared of conventional coal-fired and oil-fired power plants and of nuclear power plants. The values are compared of SO2, NO2, ash and soot emmisions with 133Xe and 85Kr fission products release and the requirement for air for diluting these emissions in the atmosphere is assessed. Also compared are thermal pollution from an oil-fired power plant and from PWR and fast reactor power plants. The conclusion is arrived at that nuclear energy can solve the problem of increasing demand for electric and heat power while reducing negative environmental impacts. (O.K.)

  17. Intelligent operation system for nuclear power plants

    International Nuclear Information System (INIS)

    Nuclear power plants consist of many systems and are operated by skillful operators with plenty of knowledge and experience of nuclear plants. Recently, plant automation or computerized operator support systems have come to be utilized, but the synthetic judgment of plant operation and management remains as human roles. Toshiba is of the opinion that the activities (planning, operation and maintenance) should be integrated, and man-machine interface should be human-friendly. We have begun to develop the intelligent operation system aiming at reducing the operator's role within the fundamental judgment through the use of artificial intelligence. (author)

  18. Validation of a methodology for the study of generation cost of electric power for nuclear power plants

    International Nuclear Information System (INIS)

    It was developed a model for the calculation of costs of electric generation of nuclear plants. The developed pattern was validated with the one used by the United States Council for Energy Awareness (USCEA) and the Electric Power Research Institute (EPRI), in studies of comparison of alternatives for electric generation of nuclear plants and fossil plants with base of gas and of coal in the United States described in the guides calls Technical Assessment Guides of EPRI. They are mentioned in qualitative form some changes in the technology of nucleo electric generation that could be included in the annual publication of Costs and Parameters of Reference for the Formulation of Projects of Investment in the Electric Sector of the Federal Commission of Electricity. These changes are in relation to the advances in the technology, in the licensing, in the construction and in the operation of the reactors called advanced as the A BWR built recently in Japan. (Author)

  19. Fault simulator trainer for nuclear power plants

    International Nuclear Information System (INIS)

    A new nuclear power plant control simulator, developed at Karlsruhe for training operators, is described, based on an office type minicomputer with visual displays representing the various relevant reactor parameters, commands, controls status and safety arrangements. (S.R.)

  20. Industrial accidents in nuclear power plants

    International Nuclear Information System (INIS)

    In 12 nuclear power plants in the Federal Republic of Germany with a total of 3678 employees, 25 notifiable company personnel accidents and 46 notifiable outside personnel accidents were reported for an 18-month period. (orig./HP)

  1. Fiber optic applications in nuclear power plants

    International Nuclear Information System (INIS)

    Fiber optic technology possesses many desirable attributes for applications in commercial nuclear power plants. The non-electrical nature of fiber optics is an important factor in an industry governed by federal safety regulations such as Class 1E isolation and separation criteria. Immunity from Electromagnetic Interference (EMI), an increasing industry problem area, is another significant characteristic. Because of the extremely wide bandwidth offered, fiber optics better addresses the data acquisition and communication requirements of the complex processes of a nuclear power plant. Potential for fiber optic sensor applications exists within the nuclear industry because their small size and physical flexibility allows access into normally inaccessible areas. They possess high accuracy and allow environmentally sensitive electronics to be remotely located. The purpose of this paper is to explore current applications for fiber optic technology in modern nuclear plants, document examples of present day usage in C-E plants and suggest possible future application areas

  2. Design quality assurance for nuclear power plants

    International Nuclear Information System (INIS)

    This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig

  3. Training program for nuclear power plant personnel

    International Nuclear Information System (INIS)

    Nuclear power generation in Japan reached 24.7% of its electric power supply with its capacity and time availability factors of 76.2% and 77.1%, respectively (in the calendar year 1986 - as of December 31, 1986). One of the reasons for such high performance is attributable to high quality of operating and maintenance personnel in the nuclear power plants. Ministry of International Trade and Industry of the Japanese Government has an overall responsibility with relation to the safety regulations and supervises all scope of training, while the Thermal and Nuclear Power Engineering Society is authorized to conduct licensing activities to qualify the chief shift supervisor of nuclear power plant operation and individual utility companies are required to train their plant operating and maintenance personnel. General status of training for plant personnel is briefly described in this paper, touching the practical education and training systems of utility companies and operation and maintenance training facilities

  4. Environmental hazards from nuclear power plants

    International Nuclear Information System (INIS)

    The article discusses the radiation exposure due to nuclear power stations in normal operation and after reactor incidents. Also mentioned is the radiation exposure to the emissions from fuel reprocessing plants and radioactive waste facilities. (RW/AK)

  5. 76 FR 1469 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

    Science.gov (United States)

    2011-01-10

    ... COMMISSION Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2... Cliffs Nuclear Power Plant, LLC, the licensee, for operation of the Calvert Cliffs Nuclear Power Plant... for light-water nuclear power reactors,'' which requires that the calculated emergency core...

  6. Feedwater system of nuclear power plant

    International Nuclear Information System (INIS)

    The present invention concerns a feedwater system of a BWR power plant and, especially, it relates to a heater drain system from a high pressure feedwater heater. That is, the feedwater system comprises (1) a feedwater pump for supplying condensates discharged from a turbine to the reactor as feedwater, (2) a feedwater heater for heating the feedwater by steams extracted from the turbine, (3) a drain reservoir for recovering the drain from the feedwater heater, (4) a drain pump for sending the drain in the drain reservoir to the upstream of the feedwater pump and (5) a drain pipeline. In this feedwater system, a drain cooling device is disposed between the drain reservoir and a drain pump, to cool the drain by the condensates in the upstream of the feedwater heater. With such a constitution, it is possible to lower the temperature of the drain on the suction side of the drain pump, which is at a temperature near the almost at a saturated temperature and tends to boil and generate bubbles. Accordingly, damages of pumps due to boiling of the drain (flush) can be prevented. (I.S.)

  7. Environmental and security challenges of nuclear plants

    International Nuclear Information System (INIS)

    The world population increase, the acceleration of global requirement for development and the need to expand energy production, have led to the depletion of natural resources. The international efforts are increasing to get clean, safe and economical energy sources . The electricity generated from nuclear energy considers less polluting and high economic competitiveness as well as reliability and efficiency. The nuclear power plants projects face significant challenges, especially after two major accidents, in Chernobyl 1986 and Fukushima 2011 including the fears of radiation effects, nuclear waste management and nuclear proliferation issues, as well as the lack of public acceptance. So those bodies interested in operating nuclear power plants work to increase nuclear safety standards, review the nuclear facilities safety, know the strict application of laws, seek to prove the economic competitiveness, maintain environmental security, assist in the nonproliferation regime and gain public acceptance. This article discusses the most important environmental and security challenges of nuclear power plants. It highlights the importance of the peaceful uses of nuclear energy as a source of sustainable development and environmental security. It also offers a number of recommendations to support the Arab countries trend towards the inclusion of nuclear energy option within their national programs to generate electricity. (author)

  8. Bivariate empirical mode decomposition applied to the estimation of out-of-phase oscillations in BWR

    International Nuclear Information System (INIS)

    Highlights: • Bivariate empirical mode decomposition (BEMD) in BWR’s instabilities is studied. • The phase determines out-of-phase oscillations in the BWR instability. • The method based in BEMD does not represent a high computational complexity. • The methodology was validated with Nuclear Power Plants stability benchmarks. • The results show that the method contributes to detect out-of-phase oscillations. - Abstract: In this paper a new method based on the bivariate empirical mode decomposition to estimate the phase of regional (out-of-phase) or global (in-phase) modes associated with instabilities in boiling water reactors (BWR), is explored. The proposed method allows decomposing the analyzed signal (constructed from two different Local Power Range Monitors, LPRMs) in different levels or intrinsic mode functions (IMF). The estimation of the phase between these LPRM signals can be achieved by tracking the modes associated to the instability of the BWR and obtaining the cross-correlation function of their corresponding IMF. This phase determines possible out-of-phase oscillations, which play an important role in the BWR instability. The method is relatively simple to implement and it does not represent a high computational complexity. The methodology was tested with simulated signals and validated with two events reported in the Forsmark and Ringhals stability benchmarks. The results of the cases studied show that the proposed method clearly contributes on the fact to detect possible cases of out-of-phase oscillations

  9. BWR stability analysis with three-dimensional transient code

    International Nuclear Information System (INIS)

    Recently, neutron flux oscillations of two different modes were observed in several foreign BWR plants. One is core wide oscillation mode which is characterized by a phenomenon that neutron flux oscillates in-phase over a whole core. At La Salle 2 plant (U.S.A.), the amplitude of core wide neutron flux oscillation grew considerably large to result in a reactor scram, which aroused great concern about BWR stability. The other is regional oscillation mode which is characterized by the phenomenon, as typically observed at Caorso plant (Italy), that neutron flux of a half core oscillates out-of-phase to that of the other half core. These neutron flux oscillation phenomena were caused by nuclear-thermal hydraulic coupled instability and requires an evaluation study on oscillation detectability and effect on fuel integrity. Particularly, the regional oscillation mode requires three-dimensional analysis since it may bring about locally large amplitude power oscillation. For this reason, analysis was done with the three-dimensional transient code TOSDYN-2 to study reactor condition which causes the regional oscillation and also to evaluate fuel thermal margin under the neutron flux oscillations of these two instability modes. (author)

  10. Report concerning Zarnowiec nuclear power plant

    International Nuclear Information System (INIS)

    Report of the Team of the President of the National Atomic Energy Agency regarding Zarnowiec nuclear power plant contains the analysis of situation in Poland in June 1990, the assessment of public opinion, as well as the description of ecological, technical and economical problems. The team's conclusions are given together with the general conclusion to stop the construction of Zarnowiec nuclear power plant. 5 appendixes, 6 enclosures, 1 documents list, 1 tab. (A.S.)

  11. Nuclear power plants in Germany. Performance 2011

    International Nuclear Information System (INIS)

    The report on the performance nuclear power plants in Germany in 2011 includes the operational results, safety analyses, revisions, quality management, environmental management, status of radioactive waste management, and eventual programs (TACIS, WANO) for the nuclear power plants Biblis A, Biblis B, Brokdorf KBR, Brunsbuettel KKB, Emsland KKE, Grafenrheinfeld KKG, Grohnde KWG, Gundremmingen KRB B, Gundremmingen KRB C, Isa KK1, Isar KK2, Kruemmel KKK, Neckarwestheim GKN I, Neckarwestheim GKN II, Philippsburg KKP 1, Philippsburg KKP 2, and Unterweser KKU.

  12. Seismic instrumentation for nuclear power plants

    International Nuclear Information System (INIS)

    A seismic instrumentation system used in Nuclear Power Plants to monitor the design parameters of systems, structures and components, needed to provide safety to those plants, against the action of earth quarks is described. The instrumentation is based on the nuclear standards and other components used, as well as their general localization is indicated. The operation of the instrumentation system as a whole and the handling of the recovered data are dealt with accordingly. The accelerometer is described in detail. (Author)

  13. Sensor network for nuclear power plant

    International Nuclear Information System (INIS)

    This study proposes an innovative method for the monitoring the nuclear power plant. In this field, false detection of the trouble, both 'false negative' and 'false positive' will become a serious problem. In the other hand, since its complexity of nuclear power plant, simplicity of monitoring system is strongly required. Here, we propose new method of signal processing and data transmission to realize reliable monitoring system consisted of multiple nodes of sensors. (author)

  14. Quality assurance organization for nuclear power plants

    International Nuclear Information System (INIS)

    This Safety Guide provides requirements, recommendations and illustrative examples for structuring, staffing and documenting the organizations that perform activities affecting quality of a nuclear power plant. It also provides guidance on control of organization interfaces, and establishment of lines for direction, communication and co-ordination. The provisions of this Guide are applicable to all organizations participating in any of the constituent areas of activities affecting quality of a nuclear power plant, such as design, manufacture, construction, commissioning and operation

  15. Problems facing a first nuclear power plant

    International Nuclear Information System (INIS)

    Requirement of nuclear power generation. Reason for considering a nuclear power programme. Decision to 'go nuclear'. Existing antecedents in the country (nuclear research institution, conventional generating plants, other nuclear utilities). - First organizational steps. Feasibility studies. Site selection and power module. Eventual reactor type decision. Site approval. - Pre-purchasing activities. Eventual selection of a consultant. Domestic participation capabilities. Pre-qualification bids. - Definition of contract type and scopes. Turn-key/non-turn-key. Architect Engineer organization. Bidding documentation. Financing. Warranties. Role of the owner. Licensing procedures and regulations. (orig./GL)

  16. Total-system expertise in economically efficient operation of nuclear power plants

    International Nuclear Information System (INIS)

    Siemens Nuclear Power GmbH can look back on well over 40 years of experience in developing and constructing nuclear power plants. 23 Power plant units of Siemens design are in operation in five countries, and in autumn this year, another one will start commercial operation, while yet another one is under construction. In comparative international power plant surveys, the Siemens-design systems usually rank in top positions when it comes to comparing systems availability and electric power generation, and Siemens have build a reputation in manufacturing power plants up to the highest safety standards worldwide. Our experience as a manufacturer of turnkey PWR and BWR type reactors, as well as our profound knowledge of international nuclear standardisation, engineering codes and safety guides, has been used and processed to the benefit of the services offered by Siemens, resulting in well-devised service packages, and enhancements and optimisation of our machinery and equipment. Siemens has of course obtained the relevant licenses and certification for all its services and products according to DIN ISO 9001, KTA and ASME standards

  17. Preliminary site decision for Geesthacht-Kruemmel nuclear power plant confirmed with reservations

    International Nuclear Information System (INIS)

    The Superior Administrative Court of the Countries of Lower Saxony and Schleswig-Holstein in Lueneburg changed the judgement of 28.5. by Schleswig-Holstein administrative court because of objection by the plaintiff. The location for the construction of a nuclear power plant with BWR reactor with a thermal power of 3 690 nW in Geesthacht-Kruemmel, right Elbe shore, loc. Kruemmel/Gruenhof, Flur 1/2, is only suitable with following preconditions: 1) The possiblity of synergistic effects in the densely populated Hamburg area must be taken in consideration thus designing the plant for doses significantly below the values mentioned in paragraph 28. pr. 3 sentences 1 and 2 of the ratiation protection law of 13 Oct 1976: 2) The escape rates of radio iodin must not exceed a value which must be determined with regard to the preexposure of areas important for milk production around Hamburg by other nuclear power plants. 3) Before starting up the nuclear power plant, it must be proven that there is no danger for the health of Hamburg's inhabitants by pollution of drinking-water with long-live radionuclides enriched during the ground passage of the Elbe water. (orig./HP)

  18. New regulatory standards and severe accident countermeasures at nuclear power plants

    International Nuclear Information System (INIS)

    Following the accident of Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Station, the basic policy for the cause for the escalation of severe accidents was shown, and based on this, new regulatory standards were enacted in July 2013. In addition to the above countermeasures, these standards intensified countermeasures for natural disasters such as tornado and volcanic eruption, as well as for fire and flooding inside power plants, and aircraft terrorism. The standards demanded that even if a fire has occurred, the cooling and monitoring of a nuclear power plant should be continued, using a designated severe accident countermeasure facility. For PWR, the installation of filter vents and designated severe accident countermeasure facility are given a grace period of five years. For BWR, measures for the breakage of containment, such as filter vents, and emission prevention measures for radioactive substances are prerequisites for the resumption of the operation. As specific examples of tsunami countermeasures of power plants across Japan, the following are introduced: (1) seawalls and tide gates at Chubu Electric Power Hamaoka Power Station, (2) seawalls at Chubu Electric Power Shimane Power Station, and (3) watertight structure of the reactor building at Tokyo Electric Power Kashiwazaki Power Station. As the severe accident countermeasures of each nuclear power plant in Japan, points are explained on the basis of the core damage prevention measures at the Units 1 and 2 of Kyushu Electric Power Sendai Power Station that are described in the license application for change of reactor installation. (A.O.)

  19. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

  20. Does Brazil need new nuclear power plants?

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Joaquim F. de [Graduate Program on Energy, University of Sao Paulo, SP (Brazil)], E-mail: jfdc35@uol.com.br; Sauer, Ildo L. [Graduate Program on Energy, University of Sao Paulo, SP (Brazil); Institute of Electrotechnics and Energy, University of Sao Paulo, SP (Brazil)], E-mail: illsauer@iee.usp.br

    2009-04-15

    In October 2008, the Brazilian Government announced plans to invest US$212 billion in the construction of nuclear power plants, totaling a joint capacity of 60,000 MW. Apart from this program, officials had already announced the completion of the construction of the nuclear plant Angra III; the construction of large-scale hydroelectric plans in the Amazon and the implantation of natural gas, biomass and coal thermoelectric plants in other regions throughout the country. Each of these projects has its proponents and its opponents, who bring forth concerns and create heated debates in the specialized forums. In this article, some of these concerns are explained, especially under the perspective of the comparative analysis of costs involved. Under such merit figures, the nuclear option, when compared to hydro plants, combined with conventional thermal and biomass-fueled plants, and even wind, to expand Brazilian power-generation capacity, does not appear as a priority.

  1. Does Brazil need new nuclear power plants?

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, Joaquim F. [Graduate Program on Energy, University of Sao Paulo, SP (Brazil); Sauer, Ildo L. [Graduate Program on Energy, University of Sao Paulo, SP (Brazil)]|[Institute of Electrotechnics and Energy, University of Sao Paulo, SP (Brazil)

    2009-04-15

    In October 2008, the Brazilian Government announced plans to invest US$212 billion in the construction of nuclear power plants, totaling a joint capacity of 60,000 MW. Apart from this program, officials had already announced the completion of the construction of the nuclear plant Angra III; the construction of large-scale hydroelectric plans in the Amazon and the implantation of natural gas, biomass and coal thermoelectric plants in other regions throughout the country. Each of these projects has its proponents and its opponents, who bring forth concerns and create heated debates in the specialized forums. In this article, some of these concerns are explained, especially under the perspective of the comparative analysis of costs involved. Under such merit figures, the nuclear option, when compared to hydro plants, combined with conventional thermal and biomass-fueled plants, and even wind, to expand Brazilian power-generation capacity, does not appear as a priority. (author)

  2. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  3. 78 FR 50458 - Entergy Nuclear Operations, Inc., James A. Fitzpatrick Nuclear Power Plant, Vermont Yankee...

    Science.gov (United States)

    2013-08-19

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Entergy Nuclear Operations, Inc., James A. Fitzpatrick Nuclear Power Plant, Vermont Yankee Nuclear Power Station, Pilgrim Nuclear Power Station, Request for Action AGENCY: Nuclear Regulatory...

  4. Current status of life management policies for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Because of growing public interest and concern on increasing aged nuclear plants in near future, the importance of measures against the aging nuclear plants was pointed out. In April 1996, the Ministry of International Trade and Industry (MITI) published the first report regarding the measures to cope with the aged nuclear power plants. The report summarizes the results of studies of technical evaluation of the aged nuclear power plants and the measures to address the aged plants. The power plants evaluated are two BWR units and one PWR unit. The first phase evaluations were focused on the major components and structures such as the reactor pressure vessels and core internals which are important for safety and are not easily repaired or replaced (Part 1 Evaluation). In the second phase, utility companies have carried out the technical evaluations not only for the major components and structures but also for all the components and structures of the plants, and the results are now under review by the government (Part 2 Evaluation). In the report, the technical evaluation concluded that with correct and adequate maintenance, safe operation is possible despite operation having exceeded 30 years. Regarding the measures, the direction to enhancement of periodical inspections and establishment of structural standards in response to the plant aging was indicated. Focused on the maintenance activities by the utilities, the report also indicates that it is important for the utilities to establish the appropriate long-term maintenance program. The report also indicated the technology development items toward attaining further highly reliable management. (author)

  5. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    A relational database management system with a query language, Reference 1, has been used to develop a Boiling Water Reactor (BWR) fuel channel tracking system on a microcomputer. The software system developed implements channel vendor and Nuclear Regulatory Commission recommendations for in-core channel movements between reactor operating cycles. A BWR Fuel channel encloses the fuel bundle and is typically fabricated using Ziracoly-4. The channel serves three functions: (1) it provides a barrier to separate two parallel flow paths, one inside the fuel assembly and the other in the bypass region outside the fuel assembly and between channels; (2) it guides the control rod as it moves between fuel assemblies and provides a bearing surface for the blades; and (3) it provides rigidity for the fuel bundle. All of these functions are necessary in typical BWR core designs. Fuel channels are not part of typical Pressurized Water Reactor (PWR) core designs

  6. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    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)

  7. Questions and Answers About Nuclear Power Plants.

    Science.gov (United States)

    Environmental Protection Agency, Washington, DC.

    This pamphlet is designed to answer many of the questions that have arisen about nuclear power plants and the environment. It is organized into a question and answer format, with the questions taken from those most often asked by the public. Topics include regulation of nuclear power sources, potential dangers to people's health, whether nuclear…

  8. Radiation protection in nuclear power plants

    International Nuclear Information System (INIS)

    The organization of workers' protection in a nuclear power plant is stated. Considering the nature and magnitude of potential risks and protection procedures, an inventory of occupational safety is made, taking account of accident statistics. It shows the credit to nuclear energy can be granted as to occupational safety

  9. Selection procedures for nuclear power plant personnel

    International Nuclear Information System (INIS)

    Selection procedures in reference to experience in staffing two Wisconsin Electric Power Company nuclear project offices and the Point Beach Nuclear Power Plant are discussed. Wisconsin Electric has had a great deal of experience in the application of psychological tests and evaluation procedures, and it was natural that a major consideration in staffing these facilities was the selection of testing procedures

  10. MODELLING OF NUCLEAR POWER PLANT DECOMMISSIONING FINANCING

    Czech Academy of Sciences Publication Activity Database

    Bemš, J.; Knápek, J.; Králík, T.; Hejhal, M.; Kubančák, Ján; Vašíček, J.

    2015-01-01

    Roč. 164, č. 4 (2015), s. 519-522. ISSN 0144-8420 Institutional support: RVO:61389005 Keywords : nuclear power plant * methodology * future decommissioning costs Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.913, year: 2014

  11. Modelling of nuclear power plant decommissioning financing

    Czech Academy of Sciences Publication Activity Database

    Bemš, J.; Knápek, J.; Králík, T.; Hejhal, M.; Kubančák, Ján; Vašíček, J.

    Vol. 2015. Oxford: Oxford Journals, 2015, s. 1-4. ISSN 1742-3406. [8th International Conference on High Levels of Natural Radiation and Radon Areas (ICHLNRRA 2014). Prague (CZ), 01.09.2014-05.09.2014] Institutional support: RVO:61389005 Keywords : nuclear power plant * methodology * future decommissioning costs Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders

  12. Effort on Nuclear Power Plants safety

    International Nuclear Information System (INIS)

    Prospects of nuclear power plant on designing, building and operation covering natural safety, technical safety, and emergency safety are discussed. Several problems and their solutions and nuclear energy operation in developing countries especially control and permission are also discussed. (author tr.)

  13. Nuclear power plant security assessment technical manual.

    Energy Technology Data Exchange (ETDEWEB)

    O' Connor, Sharon L.; Whitehead, Donnie Wayne; Potter, Claude S., III

    2007-09-01

    This report (Nuclear Power Plant Security Assessment Technical Manual) is a revision to NUREG/CR-1345 (Nuclear Power Plant Design Concepts for Sabotage Protection) that was published in January 1981. It provides conceptual and specific technical guidance for U.S. Nuclear Regulatory Commission nuclear power plant design certification and combined operating license applicants as they: (1) develop the layout of a facility (i.e., how buildings are arranged on the site property and how they are arranged internally) to enhance protection against sabotage and facilitate the use of physical security features; (2) design the physical protection system to be used at the facility; and (3) analyze the effectiveness of the PPS against the design basis threat. It should be used as a technical manual in conjunction with the 'Nuclear Power Plant Security Assessment Format and Content Guide'. The opportunity to optimize physical protection in the design of a nuclear power plant is obtained when an applicant utilizes both documents when performing a security assessment. This document provides a set of best practices that incorporates knowledge gained from more than 30 years of physical protection system design and evaluation activities at Sandia National Laboratories and insights derived from U.S. Nuclear Regulatory Commission technical staff into a manual that describes a development and analysis process of physical protection systems suitable for future nuclear power plants. In addition, selected security system technologies that may be used in a physical protection system are discussed. The scope of this document is limited to the identification of a set of best practices associated with the design and evaluation of physical security at future nuclear power plants in general. As such, it does not provide specific recommendations for the design and evaluation of physical security for any specific reactor design. These best practices should be applicable to the design and

  14. Nuclear power plant security assessment technical manual

    International Nuclear Information System (INIS)

    This report (Nuclear Power Plant Security Assessment Technical Manual) is a revision to NUREG/CR-1345 (Nuclear Power Plant Design Concepts for Sabotage Protection) that was published in January 1981. It provides conceptual and specific technical guidance for U.S. Nuclear Regulatory Commission nuclear power plant design certification and combined operating license applicants as they: (1) develop the layout of a facility (i.e., how buildings are arranged on the site property and how they are arranged internally) to enhance protection against sabotage and facilitate the use of physical security features; (2) design the physical protection system to be used at the facility; and (3) analyze the effectiveness of the PPS against the design basis threat. It should be used as a technical manual in conjunction with the 'Nuclear Power Plant Security Assessment Format and Content Guide'. The opportunity to optimize physical protection in the design of a nuclear power plant is obtained when an applicant utilizes both documents when performing a security assessment. This document provides a set of best practices that incorporates knowledge gained from more than 30 years of physical protection system design and evaluation activities at Sandia National Laboratories and insights derived from U.S. Nuclear Regulatory Commission technical staff into a manual that describes a development and analysis process of physical protection systems suitable for future nuclear power plants. In addition, selected security system technologies that may be used in a physical protection system are discussed. The scope of this document is limited to the identification of a set of best practices associated with the design and evaluation of physical security at future nuclear power plants in general. As such, it does not provide specific recommendations for the design and evaluation of physical security for any specific reactor design. These best practices should be applicable to the design and

  15. Insurance risk of nuclear power plant concentrations

    International Nuclear Information System (INIS)

    The limited number of sites available in the Federal Republic of Germany for the erection of nuclear power plants has resulted in the construction of multiple nuclear generating units on a few sites, such as Biblis, Gundremmingen and Neckarwestheim. At a value invested of approximately DM 1,200/kW this corresponds to a property concentration on one site worth DM 2 - 3 billion and more. This raises the question whether a concentration of value of this magnitude does not already exceed the limits of bearable economic risks. The property risk of a nuclear power plant, as that of any other industrial plant, is a function of the property that can be destroyed in a maximum probable loss. Insurance companies subdivide plants into so-called complex areas in which fire damage or nuclear damage could spread. While in some foreign countries twin nuclear power plants are built, where the technical systems of both units are installed in one building without any physical separation, dual unit plants are built in the Federal Republic in which the complexes with a high concentration of valuable property are physically separate building units. As a result of this separation, property insurance companies have no grounds for assessing the risk and hence, the premium different from those of single unit plants. (orig.)

  16. Use of a remotely operated vehicle (submarine) for nuclear plant inspections

    International Nuclear Information System (INIS)

    This paper describes the use of a specialized remotely operated vehicle (ROV) to perform visual inspections underwater in nuclear power plant. An underwater ROV, or minisubmarine, for visual inspections has several advantages over the more traditional camera-on-a-pole techniques and can perform some inspections easily that in the past were considered too difficult to conduct remotely. Other advantages include radiation dose savings, outage critical path savings, and reduced manpower and expenses. Some of the principal advantages of the ROV is its maneuverability, accessibility, and stability. This ROV is remarkably stable and has some stationary positioning capabilities. A necessary inspection inside a boiling water reactor (BWR) steam drier would not have been practical using other techniques. Containment suppression pool inspection was once considered nearly impossible using conventional remote inspection equipment; however, the ROV has reduced this mammoth task to one of a few person-days without the risk of having a diver in the water. Inside the BWR vessel this unit can inspect everything above the top guide. Because of the radiation-resistant color camera, the ROV proved extremely helpful during fuel movement for fuel assembly identification, orientation, and core verification. Fuel pools, reactor cavities, and storage pools all have liner seam welds that the ROV can readily inspect while they are flooded without interfering with other work occurring simultaneously

  17. Psychological empowerment in French nuclear power plants

    OpenAIRE

    Fillol, Charlotte

    2011-01-01

    Since the eighties, nuclear safety has been discussed in organizational studies and constitutes nowadays a specific stream with several standpoints. Regarding the reliability of nuclear plants, the nuclear safety literature has emphasized on the crucial role of individuals and human factors. Especially, some researchers have noticed rule breaking behavior and the impact of individual self-confidence on thebehavior; but without deepening their analyses. As high self-esteem and confidence, i.e....

  18. Nuclear power plant safety in Brazil

    International Nuclear Information System (INIS)

    The Code of Practice for the Safe Operation of Nuclear Power Plants states that: 'In discharging its responsibility for public health and safety, the government should ensure that the operational safety of a nuclear reactor is subject to surveillance by a regulatory body independent of the operating organization'. In Brazil this task is being carried out by the Comissao Nacional de Energia Nuclear in accordance with the best international practice. (orig./RW)

  19. Nuclear power plant construction activity, 1986

    International Nuclear Information System (INIS)

    Cost estimates, chronological data on construction progress, and the physical characteristics of nuclear units in commercial operation and units in the construction pipeline as of December 31, 1986, are presented. This report, which is updated annually, was prepared to provide an overview of the nuclear power plant construction industry. The report contains information on the status of nuclear generating units, average construction costs and lead-times, and construction milestones for individual reactors

  20. Cutting BWR feedwater crud levels further

    International Nuclear Information System (INIS)

    Reducing iron input to the BWR primary system is an important first step in decreasing radiation fields and occupational exposure. For feedwater iron, the specified optimum concentration is 0.1-0.5 ppb, as demonstrated by the newest ''low crud'' plants operating in Japan. Recent advances in condensate filtration will achieve the levels needed for optimised water chemistry and promise great benefits. Of particular interest is a newly developed filterdemineraliser septum that separates the filtration and ion exchange functions to allow each to be specifically optimised for BWR conditions. (author)

  1. Drought prompts government to close nuclear plant

    CERN Multimedia

    2003-01-01

    "A nuclear power plant was shut down Sunday because a record drought left insufficient water to cool down the reactor. The plant supplies more than 10 percent of Romania's electricity and closure prompted fears of a price hike" (1/2 page).

  2. Operations quality assurance for nuclear power plants

    International Nuclear Information System (INIS)

    This standard covers the quality assurance of all activities concerned with the operation and maintenance of plant equipment and systems in CANDU-based nuclear power plants during the operations phase, the period between the completion of commissioning and the start of decommissioning

  3. Availability Improvement of German Nuclear Power Plants

    International Nuclear Information System (INIS)

    High availability is important for the safety and economical performance of Nuclear Power Plants (NPP). The strategy for availability improvement in a typical German PWR shall be discussed here. Key parameters for strategy development are plant design, availability of safety systems, component reliability, preventive maintenance and outage organization. Plant design, availability of safety systems and component reliability are to a greater extent given parameters that can hardly be influenced after the construction of the plant. But they set the frame for maintenance and outage organisation which have shown to have a large influence on the availability of the plant. (author)

  4. Nuclear power plants: safety management, safety culture

    International Nuclear Information System (INIS)

    The meeting on Nuclear Power Plants: Safety Management, Safety Culture held in Munich on October 30-31, 2002 has made very important contributions to the exchange of ideas and to the analysis of the current situation with respect to this topic. Eighteen technical papers were presented in which the subject was treated and put up for discussion under a variety of aspects raised by nuclear power plant operators, regulatory authorities, expert consultants, experts, scientists, and consultants. The findings elaborated at the event will greatly help to strengthen safety further in electricity generation from nuclear power at its current high level. (orig.)

  5. BWR-GALE, Radioactive Gaseous and Liquid Waste Release from BWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  6. Application of living PSA tool FT-FREE to safety management during operation and shutdown nuclear power plant

    International Nuclear Information System (INIS)

    To date, the authors have developed FT-FREE as a tool which automatically prepares fault trees (FT) from piping and instrumentation diagram (P and ID), power supply system diagrams in BWR5 plants. FT-FREE makes it possible to perform plant safety management during operation and shutdown, with the aim of operation at nuclear power plants in the future using this tool. During operation, this function evaluates of the safety systems in terms of core damage frequency (CDF) and the reliability of the BOP systems in terms of SCRAM frequency. During shutdown, it confirms the safety of the plant accompanying changes in configuration by an evaluation using CDF. The improved tool also includes a function, which makes it possible to confirm whether or not the isolation condition of the respective component on the process chart confirms to safety management measures during shutdown. (S.Y.)

  7. Linguistic control of a nuclear power plant

    International Nuclear Information System (INIS)

    A multivariable linguistic controller based on fuzzy set theory is discussed and its application to a pressurized water nuclear power plant control is illustrated by computer simulation. The nonlinear power plant simulation model has nine states, two control inputs, one disturbance input, and two outputs. Although relatively simple, the model captures the essential coupled nonlinear plant dynamics and is convenient to use for control system studies. The use of an adaptive version of the controller is also demonstrated by computer simulation

  8. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen

    International Nuclear Information System (INIS)

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  9. Development of integral seamless fittings for nuclear power plant

    International Nuclear Information System (INIS)

    It has been required to develop the integral seamless fittings to be used for CRD or ICM housing, stick elbow and high neck tee for BWR plant. Since the pipe portion and flange or elbow portion are made together from one steel billet and not connected by welding, these integral seamless fittings have many advantages such an increase of plant reliability and no need for in-service inspection. After several trial production run, Sumitomo Metal Ind., Ltd. succeeded in manufacturing the integral seamless fittings. This report describes the manufacturing process and the test results on the integral seamless fittings. (author)

  10. Operating experience from Swedish nuclear power plants, 1999

    International Nuclear Information System (INIS)

    The total generation of electricity from Swedish nuclear power plants was 70.1 TWh during 1999, which is slightly more than the mean value for the last five years. The total electricity consumption decreased by one percent, compared with 1998, to a total of 142.3 TWh, due to an unusually warm summer and autumn. The abundant supply of hydroelectric power resulted in comparatively extensive load-following operation by the nuclear plants during the year. Production losses due to low demand totalled 3.0 TWh. The closure of Barsebaeck 1 will result in a capacity reduction exceeding 4 TWh per year. The hydroelectric power production was 70 TWh, which was 6 TWh more than during a normal year, i.e. a year with average rainfall. The remaining production sources, mainly from solid fuel plants combined with district heating contributed 9 TWh. Electricity generation by means of wind power is still increasing. There are now about 470 wind power stations, which produced 0.3 TWh during the year. The total electricity generation totalled 149.8 TWh, a three percent decrease compared with 1998. The preliminary figures for export were 15.9 TWh and for import 8.4 TWh. The figures above are calculated from the preliminary production result. A comprehensive report on electric power supply and consumption in Sweden is provided in the 1999 Annual Report from the Swedish Power Association. The unit capability factor for the PWRs at Ringhals averaged 91%, while the BWRs averaged 82% mainly due to the extended outages. The BWR reactors at Forsmark averaged as much as 93%. Forsmark 1 experienced the shortest refuelling outage ever in Sweden, only 9 days and 20 hours. In May, Oskarshamn 2 passed a historical milestone - the unit produced 100 TWh since connection to the grid in 1974. The final production day for Barsebaeck 1, which had been in commercial operation since 1975, was on November 30 when a decision by the Swedish Government revoked the operating licence. Three safety-related events

  11. Nuclear power plants 1995 - a world survey

    International Nuclear Information System (INIS)

    The atw Statistics Report compiled by atw lists 428 nuclear power plants with 363 397 gross MWe in operation in 30 countries in late 1995. Another 62 units with 55 180 gross MWe were under construction in 18 countries. This adds up to a total of 490 units with an aggregate 418 577 MWe. In the course of 1995 four units in four countries started commercial operation. In the survey of electricity generation in 1995 for which no information was made available from China and Kasachstan, a total of 417 nuclear power plants were covered. In the year under review they generated an aggregate 2 282 614 GWH, which is 3.4% more than in the previous year. The highest nuclear generation again was recorded in the USA with 705 771 GWh, followed by France with 377 021 GWh. The Grohnde power station in Germany attained the maximum annual production figure of 11 359 GWh. The survey includes nine tables indicating the generating performance of each nuclear power plant, the development of electricity generation in nuclear plants, and status of nuclear power plants at the end of 1995 arranged by countries, types of reactors, and reactor manufacturers. (orig.)

  12. Natural Circulation Performance in Nuclear Power Plants

    International Nuclear Information System (INIS)

    The present paper deals with a study of natural circulation in PWR systems, The study consists of two parts: in the first one, natural circulation in experimental facilities simulating PWR plants was analyzed. This made it possible to gather a broad data base which was assumed as a reference for the subsequent part of the research. Seven Nuclear Power Plants nodalizations and additional experimental data from ''non-PWR'' facilities have been considered in the second part of the paper. Conclusions are drawn about natural circulation capabilities derived for the seven Nuclear Power Plants nodalizations and from data base pertinent to three ''non-PWR'' facilities. (author)

  13. Natural circulation performance in nuclear power plants

    International Nuclear Information System (INIS)

    The present paper deals with a study of natural circulation in PWR systems. The study consists of two parts: in the first one, natural circulation in experimental facilities simulating PWR plants was analysed. This made it possible to gather a broad data base which was assumed as a reference for the subsequent part of the research. Seven Nuclear Power Plants nodalization and additional experimental data from 'non-PWR' facilities have been considered in the second part of the paper. Conclusions are drawn about natural circulation capabilities derived for the seven Nuclear Power Plants nodalization and from data base pertinent to three 'non-PWR' facilities. (author)

  14. Nuclear power plants are getting movable

    International Nuclear Information System (INIS)

    Russia is building the first floating nuclear plant. This ship named 'Academic Lomonosov' is 144 meter long and 30 meter wide, it carries 2 compact reactors KCT-40S we find usually on ice-breakers and submarines (although the enrichment of their nuclear fuel will be lower), each one releases an electrical power of 35 MW, together they are sufficient to power a city of 200.000 inhabitants. The 'Academic Lomonosov' is scheduled to be operating in 2012 in the Kamchatka peninsula. Its cost rounds 230 millions euros and 6 similar floating plants are planned to be built in Russia till 2020. In order to comply with the non-proliferation treaty, the floating power plants will be rented to foreign operators but not sold. The Russian owner will stay in charge of the operating and maintenance works and of the complete nuclear fuel cycle. This project confirms the actual and general trend for smaller nuclear units. (A.C.)

  15. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    The Finnish nuclear power plant units Loviisa 1 and 2 as well as TVO I and II were in operation for almost the whole fourth quarter of 1991. The load factor average was 94.7 % (the whole year 90.9 %). All the events in the last annual quarter, which are classified on the International Nuclear Event Scale, were below scale/level 0. Also the events which occurred in the other quarters of the year 1991 were rated at the scale's lowest levels. Occupational radiation doses and releases of radioactive material off-site remained well below authorised limits. Only quantities of radioactive material insignificant to radiation exposure originating in nuclear power plants were detected in the samples collected in the vicinity of the nuclear power plants

  16. Academic training for nuclear power plant operators

    International Nuclear Information System (INIS)

    In view of the increasing emphasis being placed upon academic training of nuclear power plant operators, it is important that institutions of higher education develop and implement programs which will meet the educational needs of operational personnel in the nuclear industry. Two primary objectives must be satisfied by these programs if they are to be effective in meeting the needs of the industry. One objective is for academic quality. The other primary objective is for programs to address the specialized needs of the nuclear plant operator and to be relevant to the operator's job. The Center for Nuclear Studies at Memphis State University, therefore, has developed a total program for these objectives, which delivers the programs, and/or appropriate parts thereto, at ten nuclear plant sites and with other plants in the planning stage. The Center for Nuclear Studies program leads to a Bachelor of Professional Studies degree in nuclear industrial operations, which is offered through the university college of Memphis State University

  17. Nuclear power plants in Europe 1990

    International Nuclear Information System (INIS)

    In 1989 the nuclear power plants in operation in the twelve countries of the European Community produced 589.8 TWh, which is 35.6% of the total net electricity generation of 1655.8 TWh. Of the European countries, France has the highest share of nuclear power in (net) electricity generation, i.e., 74.6% and 288.7 TWh. The nuclear power plants in Belgium produced 39 TWh, thus contributing 60.8% to the total net electricity generation. In the Federal Republic of Germany, the electricity generation in nuclear power plants rose by 2.8% in 1989 over the level of 1988, attaining 149.5 TWh, which marks roughly 39.5% of the public electricity supply. In Spain, nuclear power plants produced 56.1 TWh, which means a contribution of 38.4% to the total net electricity generation. In Hungary, 13.9 TWh were converted from nuclear power into electricity in 1989, thus achieving 47.5% of the total electricity generation in the country. In Britain, the share of nuclear power in the total net electricity generation was 63.4 TWh or 21.7%, in the Netherlands 3.8 TWh or 5.4%. (orig.)

  18. Safety Assessment - Swedish Nuclear Power Plants

    International Nuclear Information System (INIS)

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs

  19. Safety Assessment - Swedish Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B. [Luleaa Univ. of Technology (Sweden)

    1996-12-31

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs.

  20. Nuclear power plant transients: where are we

    International Nuclear Information System (INIS)

    This document is in part a postconference review and summary of the American Nuclear Society sponsored Anticipated and Abnormal Plant Transients in Light Water Reactors Conference held in Jackson, Wyoming, September 26-29, 1983, and in part a reflection upon the issues of plant transients and their impact on the viability of nuclear power. This document discusses state-of-the-art knowledge, deficiencies, and future directions in the plant transients area as seen through this conference. It describes briefly what was reported in this conference, emphasizes areas where it is felt there is confidence in the nuclear industry, and also discusses where the experts did not have a consensus. Areas covered in the document include major issues in operational transients, transient management, transient events experience base, the status of the analytical tools and their capabilities, probabilistic risk assessment applications in operational transients, and human factors impact on plant transients management

  1. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    During the first quarter of 1990 the Finnish nuclear power plant units Loviisa 1 and 2 and TVO I and II were in commercial operation for most of the time. Nuclear electricity accounted for 32.5% of the total Finnish electricity production during this quarter. The load factor average of the nuclear power plant units was 99.0%. An international nuclear event scale has been introduced for the classification of nuclear power plant events according to their nuclear and radiation safety significance. The scale first undergoes about a year long trial period in several countries. on the scale, events are divided into levels from 1 to 7 of which events at Level 7 are the most serious. Furthermore, Level 0 (Below Scale) is used for events with no safety significance. All events which occurred at the Finnish nuclear power plants this quarter are classified as Level 0. Occupational radiation doses and external releases of radioactivity were considerably below authorised limits. At the Loviisa plant, a back-up emergency feedwater system independent of the plant's other systems has been introduced which offers a new, alternative means of removing residual heat from the reactor. Owing to this system, the risk of a severe accident has been further reduced. At the TVO plants, systems have been introduced by which accident sequences which lead to containment failure could be eliminated and the consequences of a potential severe accident could be mitigated. In this report, also the release of short-lived radioactive materials along the transfer route of an irradiated sample is described which occured at the FiR 1 research reactor. The amounts of radioactive materials individuals received in their bodies in connection with this event were very low

  2. Designing for nuclear power plant maintainability and operability

    International Nuclear Information System (INIS)

    Experience has shown that maintenance and operability aspects must be addressed in the design work. ABB Atom has since long an ambition of achieving optimised, overall plant designs, and efficient feedback of growing operating experience has stepwise eliminated shortcomings, and yielded better and better plant operating performances. The records of the plants of the latest design versions are very good; four units in Sweden have operated at an energy availability of 90.1%, and the two Olkiluoto units in Finland at a load factor of 92.7%, over the last decade. The occupational radiation exposures have also been at a low level. The possibilities for implementing 'lessons learned' in existing plants are obviously limited by practical constraints. In Finland and Sweden, significant modernisations are still underway, however, involving replacement of mechanical equipment, and upgrading and backfitting of I and C systems on a large scale, in most of the plants. The BWR 90 design focuses on meeting requirements from utilities as well as new regulatory requirements, with a particular emphasis on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimisation of buildings and containment to decrease construction time and costs, and selection of materials as well as maintenance of operating procedures to reduce radiation exposures even further. The BWR 90 design was offered to Finland in the early 1990s, but development work continues. It has been selected by a number of European utilities for assessing its conformance with the European Utility Requirements (EUR), aiming at a specific EUR Volume 3 for the BWR 90. Some characteristics of the ABB BWRs, with emphasis on features of importance for achieving improved economy and enhanced safety, are described below. (author)

  3. Investigation and analysis of hydrogen ignition and explosion events in foreign nuclear power plants

    International Nuclear Information System (INIS)

    Reports about hydrogen ignition and explosion events in foreign nuclear power plants from 1980 to 2001 were investigated, and 31 events were identified. Analysis showed that they were categorized in (1) outer leakage ignition events and (2) inner accumulation ignition events. The dominant event for PWR (pressurized water reactor) was outer leakage ignition in the main generator, and in BWR (boiling water reactor) it was inner accumulation ignition in the off-gas system. The outer leakage ignition was a result of work process failure with the ignition source, operator error, or main generator hydrogen leakage. The inner accumulation ignition events were caused by equipment failure or insufficient monitoring. With careful preventive measures, the factors leading to these events could be eliminated. (author)

  4. Safety provisions of nuclear power plants

    International Nuclear Information System (INIS)

    Safety of nuclear power plants is determined by a deterministic approach complemented by probabilistic considerations. Much use has been made of the wealth of information from more than 6000 years of reactor operation. Design, construction and operation is governed by national and international safety standards and practices. The IAEA has prepared a set of Nuclear Safety Standards as recommendations to its Member States, covering the areas of siting, design, operations, quality assurance, and governmental organisations. In 1988 the IAEA published a report by the International Nuclear Safety Advisory Group on Basic Safety Principles for Nuclear Power Plants, summarizing the underlying objectives and principles of excellence in nuclear safety and the way in which its aspects are interrelated. The paper will summarize some of the key safety principles and provisions, and results and uses of Probabilistic Safety Assessments. Some comments will be made on the safety of WWER 440/230 and WWER-1000 reactors which are operated on Bulgaria. 8 figs

  5. SWOT of nuclear power plant sustainable development

    International Nuclear Information System (INIS)

    SWOT Analysis is a Useful tool that can he applied to most projects or business ventures. In this article we are going to examine major strengths, weaknesses, opportunities and threats of nuclear power plants in view of sustainable development. Nuclear power plants have already attained widespread recognition for its benefits in fossil pollution abatement, near-zero green house gas emission, price stability and security of energy supply. The impressive new development is that these virtues are now a cost -free bonus, because, in long run, nuclear energy has become an inexpensive way to generate electricity. Nuclear energy's pre-eminence economically and environmentally has two implications for government policy. First, governments should ensure that nuclear licensing and safety oversight arc not only rigorous but also efficient in facilitating timely development of advanced power plants. Second, governments should be bold incentivizing the transformation to clean energy economics, recognizing that such short-term stimulus will, in the case of nuclear plants, simply accelerate desirable changes that now have their own long-term momentum. The increased competitiveness of nuclear power plant is the result of cost reductions in all aspects of nuclear economics: Construction, financing, operations, waste management and decommissioning. Among the cost-lowering factors are the evolution to standardized reactor designs, shorter construction periods, new financing techniques, more efficient generation technologies, higher rates of reactor utilization, and longer plant lifetimes. U.S World Nuclear Association report shows that total electricity costs for power plant construction and operation were calculated at two interest rates. At 10%, midrange generating costs per kilowatt-hour are nuclear at 4 cents, coal at 4.7 cents and natural gas at 5.1 cent. At a 5% interest rate, mid-range costs per KWh fall to nuclear at 2.6 cents, coal at 3.7 cents and natural gas at 4.3 cents

  6. Advanced nuclear power plants in Korea

    International Nuclear Information System (INIS)

    Full text: Korea Hydro and Nuclear Power Co., Ltd (KHNP) is the largest power company among the six subsidiaries that separated from Korea Electric Power Corporation (KEPCO) in 2001, accounting for approximately 25% of electricity producing facilities, hydro and nuclear combined. KHNP operates 20 nuclear power plants in Kori, Yonggwang, Ulchin and Wolsong site and several hydroelectric power generation facilities, providing approximately 36% of the national power supply. As a major source of electricity generation in Korea, nuclear energy contributes greatly to the stability of national electricity supply and energy security. KHNP's commercial nuclear power plant operation, which started with Kori Unit 1 in 1978, has achieved an average capacity factor more than 90% since 2000 and a high record of 93.4% in 2008. Following the introduction of nuclear power plants in the 1970's, Korea accumulated its nuclear technology in the 1980's, developed OPR 1000(Optimized Power Reactor) and demonstrated advanced level of its nuclear technology capabilities in the 2000's by developing an advanced type reactor, APR 1400(Advanced Power Reactor) which is being constructed at Shin-Kori Unit 3 and 4 for the first time. By 2022, KHNP will construct additional 12 nuclear power plants in order to ensure a stable power supply according to the Government Plan of Long-Term Electricity supply and Demand. 4 units of OPR 1000 reactor model will be commissioned by 2013 and 8 units of APR 1400 are under construction and planned. At the end of 2022, the nuclear capacity will reach 33% share of total generation capacity in Korea and account for 48% of national power generation. (author)

  7. Flex concept for US-A BWR extended loss of AC power events

    International Nuclear Information System (INIS)

    The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  8. Flex concept for US-A BWR extended loss of AC power events

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  9. An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Masashi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-07-01

    This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

  10. License renewal - an idea whose time has come. Hatch nuclear plant license renewal program: an actual example of application of the license renewal rule to the Intake Structure

    International Nuclear Information System (INIS)

    After the NRC issued a revised license renewal rule in May 1995, the nuclear industry focussed on developing generic industry for implementing the rule and testing the guidance through various demonstration programs and work products in conjunction with the NRC. In addition, plant-specific programs also proceeded forward. These activities show that implementation issues continue to exist. Since the issuance of the rule, the NRC has issued a draft standard review plan for license renewal (SRP-LR), working draft, September 1997. Southern Nuclear Operating Company (SNC) has begun development work on a license renewal application for Plant Hatch Units 1 and 2. Plant Hatch Units 1 and 2 are BWR 4, Mark I plants whose operating licenses expire in 2014 and 2018, respectively. The Plant Hatch initiative also involves teaming with other boiling water reactors (BWRs) to develop the license renewal technology within the BWR fleet, and to support Plant Hatch by providing an oversight role for the application process. The teaming effort involved two other utilities, each being assigned to prepare a common report on a mechanical system or a structure. The common report could be presented to the NRC with modifications to suit the individual plants, thereby saving time and money, and hopefully resulting in quicker approval by the NRC. The desired license renewal process end result is a renewed license with up to a 20 year extension (10CFR 54.31(b)). (orig.)

  11. 76 FR 39908 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

    Science.gov (United States)

    2011-07-07

    ... COMMISSION Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2.... DPR-53 and DPR-69, for the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (CCNPP), respectively... (ISFSI), currently held by Calvert Cliffs Nuclear Power Plant, LLC as owner and licensed...

  12. 75 FR 66802 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

    Science.gov (United States)

    2010-10-29

    ... COMMISSION Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2... Regulatory Commission (the Commission) has granted the request of Calvert Cliffs Nuclear Power Plant, LLC... Operating License Nos. DPR-53 and DPR-69 for the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and...

  13. Risks of potential accidents of nuclear power plants in Europe

    OpenAIRE

    Slaper H; Eggink GJ; Blaauboer RO

    1993-01-01

    Over 200 nuclear power plants for commercial electricity production are presently operational in Europe. The 1986 accident with the nuclear power plant in Chernobyl has shown that severe accidents with a nuclear power plant can lead to a large scale contamination of Europe. This report is focussed on an integrated assessment of probabilistic cancer mortality risks due to possible accidental releases from the European nuclear power plants. For each of the European nuclear power plants the prob...

  14. The earthquake security of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    The seismic safety of Paks Nuclear Power Plant is analyzed. Assessment of earthquake risk has been done at the site of the plant, and seismic resistance of the nuclear power plant is analyzed together. (TRA)

  15. Ground assessment methods for nuclear power plant

    International Nuclear Information System (INIS)

    It is needless to say that nuclear power plant must be constructed on the most stable and safe ground. Reliable assessment method is required for the purpose. The Ground Integrity Sub-committee of the Committee of Civil Engineering of Nuclear Power Plant started five working groups, the purpose of which is to systematize the assessment procedures including geological survey, ground examination and construction design. The works of working groups are to establishing assessment method of activities of faults, standardizing the rock classification method, standardizing assessment and indication method of ground properties, standardizing test methods and establishing the application standard for design and construction. Flow diagrams for the procedures of geological survey, for the investigation on fault activities and ground properties of area where nuclear reactor and important outdoor equipments are scheduled to construct, were established. And further, flow diagrams for applying investigated results to design and construction of plant, and for determining procedure of liquidification nature of ground etc. were also established. These systematized and standardized methods of investigation are expected to yield reliable data for assessment of construction site of nuclear power plant and lead to the safety of construction and operation in the future. In addition, the execution of these systematized and detailed preliminary investigation for determining the construction site of nuclear power plant will make much contribution for obtaining nation-wide understanding and faith for the project. (Ishimitsu, A.)

  16. Nuclear power plant common aging terminology

    International Nuclear Information System (INIS)

    This report defines and encourages the use of a set of common terms and definitions that characterize aging and aging management of nuclear power plant systems, structures, and components. Common aging terminology encompasses the areas of degradation causes, degradation/aging, life, failure, and maintenance. The terminology should prove useful in plant license renewal, life-cycle management, maintenance, and equipment qualification. The principal intended benefits are improved reporting and interpretation of plant experience data (especially failure data) and improved interpretation and compliance with codes, standards, and regulations related to aging of nuclear power plants. The terminology was developed using a sound, systematic technical and lexicographical approach, as well as input from a committee with members from electric utilities, the Nuclear Management and Resources Council (NUMARC), the Electric Power Research Institute (EPRI), the United States Nuclear Regulatory Commission (NRC), and national research laboratories. Draft definitions received an extensive industry review. Formal comments were received from 19 organizations worldwide. The report presents the results of this effort and contains definitions, figures, examples of usage, dictionary definitions, and discussions describing the relationships among the 85 terms and 27 synonyms of the common aging terminology. The terms and definitions are generally consistent with similar terminology used by the Institute of Nuclear Power Operations (INPO). The common aging terminology has been reviewed by NUMARC and the NRC and is recommended for use throughout the nuclear power industry

  17. Plant life management and maintenance technologies for nuclear power plants

    International Nuclear Information System (INIS)

    Nuclear power generation occupying an important position for energy source in Japan and supplying about one third of total electric power usage is now required for further upgrading of its economics under regulation relaxation of electric power business. And, under execution retardation of its new planning plant, it becomes important to operate the already established plants for longer term and to secure their stability. Therefore, technical development in response to the plant life elongation is promoted under cooperation of the Ministry of Economics and Industries, electric power companies, literate, and plant manufacturers. Under such conditions, the Hitachi, Ltd. has progressed some technical developments on check inspection, repairs and maintenance for succession of the already established nuclear power plants for longer term under securing of their safety and reliability. And in future, by proposing the check inspection and maintenance program combined with these technologies, it is planned to exert promotion of maintenance program with minimum total cost from a viewpoint of its plant life. Here were described on technologies exerted in the Hitachi, Ltd. such as construction of plant maintenance program in response to plant life elongation agreeing with actual condition of each plant, yearly change mechanism grasping, life evaluation on instruments and materials necessary for maintenance, adequate check inspection, repairs and exchange, and so forth. (G.K.)

  18. Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO2/year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

  19. Safety criteria for nuclear chemical plants

    International Nuclear Information System (INIS)

    Safety measures have always been required to limit the hazards due to accidental release of radioactive substances from nuclear power plants and chemical plants. The risk associated with the discharge of radioactive substances during normal operation has also to be kept acceptably low. BNFL (British Nuclear Fuels Ltd.) are developing risk criteria as targets for safe plant design and operation. The numerical values derived are compared with these criteria to see if plants are 'acceptably safe'. However, the criteria are not mandatory and may be exceeded if this can be justified. The risk assessments are subject to independent review and audit. The Nuclear Installations Inspectorate also has to pass the plants as safe. The assessment principles it uses are stated. The development of risk criteria for a multiplant site (nuclear chemical plants tend to be sited with many others which are related functionally) is discussed. This covers individual members of the general public, societal risks, risks to the workforce and external hazards. (U.K.)

  20. Life cycle planning for nuclear plant staffing

    International Nuclear Information System (INIS)

    Current designs of new nuclear power plants include operational life cycles of 60 years. When the total life cycle of a new nuclear power program is considered, and includes all aspects from plant design selection to decommissioning, the total program duration may begin to approach 100 years. One hundred years of program duration requires multiple generations of workers. This paper will describe a process for considering staffing requirements for each phase of the nuclear power plant life cycle: 1) Plant Design Selection; 2) Site Selection; 3) Plant Licensing; 4) Construction; 5) Start Up/Testing; 6) Commercial Operations; 7) Shut Down/Safe Store; and 8) Decommissioning. During each of the eight life cycle phases, specific functional activities need to be performed. Thus, each phase has unique staffing requirements that must be analyzed to ensure safe and effective implementation of all required programs and activities. This paper will discuss the staffing functions relevant to each life cycle phase, and identify key drivers during each phase that directly impact staffing requirements for each of nine functional areas: 1) Operations; 2) Maintenance; 3) Work Management; 4) Radiation Protection; 5) Safety; 6) Regulatory Compliance; 7) Engineering and Technical Services; 8) Management and Support; and 9) Supply Chain. Issues surrounding management approaches will also be discussed, including the impacts of organizational design structures, out-sourcing, and work force planning. Additionally, potential impacts resulting from centralization and standardization across multiple nuclear plant sites will be discussed. (author)