Advances in commercial heavy water reactor power stations

International Nuclear Information System (INIS)

Brooks, G.L.

1987-01-01

Generating stations employing heavy water reactors have now firmly established an enviable record for reliable, economic electricity generation. Their designers recognize, however, that further improvements are both possible and necessary to ensure that this reactor type remains attractively competitive with alternative nuclear power systems and with fossil-fuelled generation plants. This paper outlines planned development thrusts in a number of important areas, viz., capital cost reduction, advanced fuel cycles, safety, capacity factor, life extension, load following, operator aida, and personnel radiation exposure. (author)

Power distribution effects on boiling water reactor stability

International Nuclear Information System (INIS)

Damiano, B.; March-Leuba, J.

1989-01-01

The work presented in this paper deals with the effects of spatial power distributions on the stability of boiling water reactors (BWRs). It is shown that a conservative power distribution exists for which the stability is minimal. These results are relevant because they imply that bounding stability calculations are possible and, thus, a worst-possible scenario may be defined for a particular BWR geometry. These bounding calculations may, then, be used to determine the maximum expected limit-cycle peak powers

Power spectral density measurements with 252Cf for a light water moderated research reactor

International Nuclear Information System (INIS)

King, W.T.; Mihalczo, J.T.

1979-01-01

A method of determining the reactivity of far subcritical systems from neutron noise power spectral density measurements with 252 Cf has previously been tested in fast reactor critical assemblies: a mockup of the Fast Flux Test Facility reactor and a uranium metal sphere. Calculations indicated that this measurement was feasible for a pressurized water reactor (PWR). In order to evaluate the ability to perform these measurements with moderated reactors which have long prompt neutron lifetimes, measurements were performed with a small plate-type research reactor whose neutron lifetime (57 microseconds) was about a factor of three longer than that of a PWR and approx. 50% longer than that of a boiling water reactor. The results of the first measurements of power spectral densities with 252 Cf for a water moderated reactor are presented

Materialistic Aspects of Raising Resource of Pressurized Water Reactors for Low-Power Nuclear Plants

International Nuclear Information System (INIS)

Parshin, A.M.; Muratov, O.E.

2005-01-01

The opportunity of using ships reactors for low-power nuclear plants is considered. Some aspects of working constructional materials on cases of water-water reactors of ships nuclear units are considered. Advantages of raising resource of ships reactors are shown

The use of ferritic materials in light water reactor power plants

International Nuclear Information System (INIS)

Marston, T.V.

1984-01-01

This paper reviews the use of ferritic materials in LWR power plant components. The two principal types of LWR systems, the boiling water reactor (BWR) and the pressurized water reactor (PWR) are described. The evolution of the construction materials, including plates and forgings, is presented. The fabrication process for both reactors constructed with plates and forgings are described in detail. Typical mechanical properties of the reactor vessel materials are presented. Finally, one critical issue radiation embrittlement dealing with ferritic materials is discussed. This has been one of the major issues regarding the use of ferritic material in the construction of LWR pressure vessels

Reactor power control device

International Nuclear Information System (INIS)

Imaruoka, Hiromitsu.

1994-01-01

A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

Next generation light water reactors

International Nuclear Information System (INIS)

Omoto, Akira

1992-01-01

In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

International Nuclear Information System (INIS)

Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

2012-09-01

selection of chemistry controls is vital for NPPs with liquid metal cooled reactors. This paper highlights principles and approaches to chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors. The recommendations on how to arrange chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors are based taking into account: - the experience with operation of fossil power industry; - secondary side water chemistry of lead-bismuth eutectics cooled nuclear reactors at submarines; - steam/water cycles of NPPs with sodium cooled fast breeders BN-350 and BN-600; - secondary water chemistry at conventional NPPs with WER, RBMK and some other reactors. (authors)

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

The Influence of RSG-GAS Primary Pump Operation Concerning the Rise Water Level of Reactor Pool in 15 MW Reactor Power

International Nuclear Information System (INIS)

Djunaidi

2004-01-01

The expansion of air volume in the delay chamber shows in rise water level of reactor pool during the operation. The rises of water level in the reactor pool is not quite from the expansion of air volume in the delay chamber, but some influence the primary pump operation. The purpose evaluated of influence primary pump is to know the influence primary pump power concerning the rise water level during the reactor operation. From the data collection during 15 MW power operation in the last core 42 the influence of primary pump operation concerning the rise water level in the reactor pool is 34.48 % from the total increased after operation during 12 days. (author)

Topics to be covered in safety analysis reports for nuclear power plants with pressurized water reactors or boiling water reactors in the F.R.G

International Nuclear Information System (INIS)

Kohler, H.A.G.

1977-01-01

This manual aims at defining the standards to be used in Safety Analysis Reports for Nuclear Power Plants with Pressurized Water Reactors or Boiling Water Reactors in the Federal Republic of Germany. The topics to be covered are: Information about the site (geographic situation, settlement, industrial and military facilities, transport and communications, meteorological conditions, geological, hydrological and seismic conditions, radiological background), description of the power plant (building structures, safety vessel, reactor core, cooling system, ventilation systems, steam power plant, electrical facilities, systems for measurement and control), indication of operation (commissioning, operation, safety measures, radiation monitoring, organization), incident analysis (reactivity incidents, loss-of-coolant incidents, external impacts). (HP) [de

An investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a boiling water reactor anticipated transient without scram

International Nuclear Information System (INIS)

Peterson, C.E.; Chexal, V.K.; Gose, G.C.; Hentzen, R.D.; Layman, W.H.

1985-01-01

Under certain anticipated transient without scram (ATWS) sequences for a boiling water reactor, it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and by helping to keep the suppression pool temperature within limits. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce reactor power. The analyses provide an assessment of the power level that would be attained, assuming the reactor operators were to reduce the the downcomer level down to the top of the active fuel

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, W.H.

1997-01-01

Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

International Nuclear Information System (INIS)

Pinedo V, J.L.

1979-01-01

One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

Method of controlling power of a heavy water reactor

International Nuclear Information System (INIS)

Masuda, Hiroyuki.

1975-01-01

Object: To adjust a level of heavy water in a region of reflection body to control power in a heavy water reactor. Structure: The interior of a core tank filled with heavy water is divided by a partition into a core heavy water region and a reflection body region formed by surrounding the core heavy water region, and a level of heavy water within the reflection body region is adjusted to control power. Preferably, it is desirable to communicate the core heavy water region with the reflection body heavy water region at their lower portion, and gas pressure applied to an upper portion within at least one of said regions is adjusted to adjust the level of heavy water within the reflection body heavy water region. Thereby, the heavy water within the reflection body heavy water region may be introduced into the core region, thus requiring no tank which stores heavy water within the reflection body region. (Kamimura, M.)

Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

International Nuclear Information System (INIS)

Permana, S.; Takaki, N.; Sekimoto, H.

2007-01-01

Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 2 33U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

Energy Technology Data Exchange (ETDEWEB)

Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

2004-11-01

One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

Power distribution monitoring system in the boiling water cooled reactor core

International Nuclear Information System (INIS)

Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

1987-01-01

Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

Advanced light-water reactors

International Nuclear Information System (INIS)

Golay, M.W.; Todreas, N.E.

1990-01-01

Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

International Nuclear Information System (INIS)

Huber, Horacio

1989-01-01

Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, William H. Sr.

1998-01-01

Self-powered neutron detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors worldwide. This paper describes the basic properties of these radiation sensors including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs, which are being effectively, used in in-core instrumentation systems for pressurized water, heavy water and graphite moderated light water reactors. Also examples are shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurized water and heavy water reactors worldwide. (author)

Risk contribution from low power and shutdown of a pressurized water reactor

International Nuclear Information System (INIS)

Chu, T.L.; Pratt, W.T.

1997-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. 12 refs., 7 tabs

Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

International Nuclear Information System (INIS)

Ganguly, Chaitanyamoy

2015-01-01

Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

Capital cost: pressurized water reactor plant. Commercial electric power cost studies

Energy Technology Data Exchange (ETDEWEB)

1977-06-01

The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

Capital cost: pressurized water reactor plant. Commercial electric power cost studies

International Nuclear Information System (INIS)

1977-06-01

The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description

SWR 1000: The new boiling water reactor power plant concept

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

International Nuclear Information System (INIS)

De, S.K.; Karhadkar, C.G.

2017-01-01

An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

Power generation versus fuel production in light water hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1977-06-01

The economic potentials of fissile-fuel-producing light-water hybrid reactors (FFP-LWHR) and of fuel-self-sufficient (FSS) LWHR's are compared. A simple economic model is constructed that gives the capital investment allowed for the hybrid reactor so that the cost of electricity generated in the hybrid based energy system equals the cost of electricity generated in LWR's. The power systems considered are LWR, FSS-LWHR, and FFP-LWHR plus LWR, both with and without plutonium recycling. The economic potential of FFP-LWHR's is found superior to that of FSS-LWHR's. Moreover, LWHR's may compete, economically, with LWR's. Criteria for determining the more economical approach to hybrid fuel or power production are derived for blankets having a linear dependence between F and M. The examples considered favor the power generation rather than fuel production

Method of operating heavy water moderated reactors

International Nuclear Information System (INIS)

Masuda, Hiroyuki.

1980-01-01

Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

Light-water nuclear reactors

International Nuclear Information System (INIS)

Drevon, G.

1983-01-01

This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

Energy Technology Data Exchange (ETDEWEB)

Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

2017-02-15

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Power reactor noise

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

International Nuclear Information System (INIS)

Nieder, R.; Resch, G.

1976-01-01

The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Reactor water level control device

International Nuclear Information System (INIS)

Utagawa, Kazuyuki.

1993-01-01

A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

Supercritical Water Reactor Cycle for Medium Power Applications

International Nuclear Information System (INIS)

BD Middleton; J Buongiorno

2007-01-01

Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency (ge)20%; Steam turbine outlet quality (ge)90%; and Pumping power (le)2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

Real-time stability monitoring method for boiling water reactor nuclear power plants

International Nuclear Information System (INIS)

Fukunishi, K.; Suzuki, S.

1987-01-01

A method for real-time stability monitoring is developed for supervising the steady-state operation of a boiling water reactor core. The decay ratio of the reactor power fluctuation is determined by measuring only the output neutron noise. The concept of an inverse system is introduced to identify the dynamic characteristics of the reactor core. The adoption of an adaptive digital filter is useful in real-time identification. A feasibility test that used measured output noise as an indication of reactor power suggests that this method is useful in a real-time stability monitoring system. Using this method, the tedious and difficult work for modeling reactor core dynamics can be reduced. The method employs a simple algorithm that eliminates the need for stochastic computation, thus making the method suitable for real-time computation with a simple microprocessor. In addition, there is no need to disturb the reactor core during operation. Real-time stability monitoring using the proposed algorithm may allow operation under less stable margins

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

International Nuclear Information System (INIS)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

Energy Technology Data Exchange (ETDEWEB)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

International Nuclear Information System (INIS)

Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

2009-01-01

R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

Capital cost: pressurized water reactor plant. Commerical electric power cost studies

International Nuclear Information System (INIS)

1977-06-01

The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate

The low-temperature water-water reactor for district heating atomic power plant (DHPP)

International Nuclear Information System (INIS)

Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Yu.V.

1977-01-01

The district heating atomic power plant in the article is distinguished by the increased reliability and safety of operation that was provided by the use of following main principles: relatively low parameters of the coolant; the intergral arrangement of equipment and accordingly the minimum branching of the reactor circuit; the natural circulation of coolant of the primary circuit in the steady-state, transient and emergency regimes of reactor operation; the considerable reserves of cold water of the primary circuit in the reactor vessel, providing the emergency cooling; the application of two shells each of which is designed for the total working pressure, the second shell is made of prestressed reinforced concrete that eliminates its brittle failure. (M.S.)

The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

International Nuclear Information System (INIS)

Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

2012-01-01

This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

On-line test of power distribution prediction system for boiling water reactors

International Nuclear Information System (INIS)

Nishizawa, Y.; Kiguchi, T.; Kobayashi, S.; Takumi, K.; Tanaka, H.; Tsutsumi, R.; Yokomi, M.

1982-01-01

A power distribution prediction system for boiling water reactors has been developed and its on-line performance test has proceeded at an operating commercial reactor. This system predicts the power distribution or thermal margin in advance of control rod operations and core flow rate change. This system consists of an on-line computer system, an operator's console with a color cathode-ray tube, and plant data input devices. The main functions of this system are present power distribution monitoring, power distribution prediction, and power-up trajectory prediction. The calculation method is based on a simplified nuclear thermal-hydraulic calculation, which is combined with a method of model identification to the actual reactor core state. It has been ascertained by the on-line test that the predicted power distribution (readings of traversing in-core probe) agrees with the measured data within 6% root-mean-square. The computing time required for one prediction calculation step is less than or equal to 1.5 min by an HIDIC-80 on-line computer

Power density effect on feasibility of water cooled thorium breeder reactor

International Nuclear Information System (INIS)

Sidik, Permana; Takaki, Naoyuki; Sekimoto, Hiroshi

2008-01-01

Breeding is made possible by the high value of neutron regeneration ratio η for 233 U in thermal energy region. The reactor is fueled by 233 U-Th oxide and it has used the light water as moderator. Some characteristics such as spectrum, η value, criticality, breeding performance and number density are evaluated. Several power densities are evaluated in order to analyze its effect to the breeding performance. The η value of fissile 233 U obtains higher value than 2 which may satisfy the breeding capability especially for thermal reactor for all investigated MFR. The increasing enrichment and decreasing conversion ratio are more significant for MFR 233 U enrichment. Number density of 233 Pa decreases significantly with decreasing power density which leads the reactor has better breeding performance because lower capture rate of 233 Pa. (author)

Nuclear power/water pumping-up composite power plant

International Nuclear Information System (INIS)

Okamura, Kiyoshi.

1995-01-01

In a nuclear power/water pumping-up composite power plant, a reversible pump for pumping-up power generation connected to a steam turbine is connected to an upper water reservoir and a lower water reservoir. A pumping-up steam turbine for driving the turbine power generator, a hydraulic pump for driving water power generator by water flowing from the upper water reservoir and a steam turbine for driving the pumping-up pump by steams from a nuclear reactor are disposed. When power demand is small during night, the steam turbine is rotated by steams of the reactor, to pump up the water in the lower water reservoir to the upper water reservoir by the reversible pump. Upon peak of power demand during day time, power is generated by the steams of the reactor, as well as the reversible pump is rotated by the flowing water from the upper water reservoir to conduct hydraulic power generation. Alternatively, hydraulic power generation is conducted by flowing water from the upper reservoir. Since the number of energy conversion steps in the combination of nuclear power generation and pumping-up power generation is reduced, energy loss is reduced and utilization efficiency can be improved. (N.H.)

Robust feedback-linearization control for axial power distribution in pressurized water reactors during load-following operation

OpenAIRE

M. Zaidabadi nejad; G.R. Ansarifar

2018-01-01

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the impor...

Procedure for operating a heavy water cooled power reactor

International Nuclear Information System (INIS)

Rau, P.; Kumpf, H.

1981-01-01

Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1978-01-01

Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

Calibration of RB reactor power

International Nuclear Information System (INIS)

Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

1976-09-01

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

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

International Nuclear Information System (INIS)

1986-09-01

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

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Current status of nuclear power generation in Japan and directions in water cooled reactor technology development

International Nuclear Information System (INIS)

Miwa, T.

1991-01-01

Electric power demand aspects and current status of nuclear power generation in Japan are outlined. Although the future plan for nuclear power generation has not been determined yet the Japanese nuclear research centers and institutes are investigating and developing some projects on the next generation of light water reactors and other types of reactors. The paper describes these main activities

Development of the fuel-cycle costs in nuclear power stations with light-water reactors

International Nuclear Information System (INIS)

Brosch, R.; Moraw, G.; Musil, G.; Schneeberger, M.

1976-01-01

The authors investigate the fuel-cycle costs in nuclear power stations with light-water reactors in the Federal Republic of Germany in the years 1966 to 1976. They determine the effect of the price development for the individual components of the nuclear fuel cycle on the fuel-cycle costs averaged over the whole power station life. Here account is taken also of inflation rates and the change in the DM/US $ parity. In addition they give the percentage apportionment of the fuel-cycle costs. The authors show that real fuel-cycle costs for nuclear power stations with light-water reactors in the Federal Republic of Germany have risen by 11% between 1966 and 1976. This contradicts the often repeated reproach that fuel costs in nuclear power stations are rising very steeply and are no longer competitive. (orig.) [de

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

International Nuclear Information System (INIS)

Reid, Robert S.; Pearson, J. Bosie; Stewart, Eric T.

2007-01-01

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

International Nuclear Information System (INIS)

Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

2007-01-01

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

Energy Technology Data Exchange (ETDEWEB)

REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

2007-01-16

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Advances in light water reactor technologies

CERN Document Server

Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

2010-01-01

""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

A potential of boiling water power reactors with a natural circulation of a coolant

International Nuclear Information System (INIS)

Osmachkin, V.S.; Sokolov, I.N.

1998-01-01

The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

Power ramping test in the JMTR for PCI study of water reactor fuel

International Nuclear Information System (INIS)

Nakata, H.; Kanbara, M.; Ichikawa, M.

1984-01-01

Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

Reactor water level control device

International Nuclear Information System (INIS)

Hiramatsu, Yohei.

1980-01-01

Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

Power ramping, cycling and load following behaviour of water reactor fuel

International Nuclear Information System (INIS)

1988-05-01

The present meeting was scheduled by the International Atomic Energy Agency upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. Sixty-three participants representing 15 countries and one international organization attended the meeting. Twenty papers were presented during three technical sessions, followed by panel discussions which allowed to formulate the conclusions of the meeting and recommendations to the Agency. The objective of this Technical Committee Meeting is to review the ''State-of-the-Art'', make critical comments and recommendations with the aim of improving fuel reliability and assure integrity of the cladding and core materials when subjected to ramping and cycling sequences. The Meeting was organized in three sessions: Session 1. ''Mechanical Behaviour and Fission Gas Release'' (7 papers); Session 2. ''Power Ramping and Power Cycling Demonstration Programmes in Research Reactors'' (5 papers); Session 3. ''Fuel Behaviour in Power Reactors'' (9 papers). Between the sessions, the session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report. A separate abstract was prepared for each of these 21 presentations. Refs, figs and tabs

The Optimization of power reactor control system

International Nuclear Information System (INIS)

Danupoyo, S.D.

1997-01-01

A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

Status of advanced technology and design for water cooled reactors: Light water reactors

International Nuclear Information System (INIS)

1988-10-01

Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

Reactor power control method and device

International Nuclear Information System (INIS)

Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

1997-01-01

The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Corrosion control in CANDU nuclear power reactors

International Nuclear Information System (INIS)

Lesurf, J.E.

1974-01-01

Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

HOMOGENEOUS NUCLEAR POWER REACTOR

Science.gov (United States)

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Introduction to the neutron kinetics of nuclear power reactors

CERN Document Server

Tyror, J G; Grant, P J

2013-01-01

An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

Energy Research Advisory Board, Civilian Nuclear Power Panel: Subpanel 1 report, Light water reactor utilization and improvement: Volume 2

International Nuclear Information System (INIS)

1986-10-01

The Secretary of Energy requested that the Office of Nuclear Energy prepare a strategic national plan that outlines the Department's role in the future development of civilian nuclear power and that the Energy Research Advisory Board establish an ad hoc panel to review and comment on this plan. The Energy Research Advisory Board formed a panel for this review and three subpanels were formed. One subpanel was formed to address the institutional issues surrounding nuclear power, one on research and development for advanced nuclear power plants and a third subpanel on light water reactor utilization and improvement. The subpanel on light water reactors held two meetings at which representatives of the DOE, the NRC, EPRI, industry and academic groups made presentations. This is the report of the subpanel on light water reactor utilization and improvement. This report presents the subpanel's assessment of initiatives which the Department of Energy should undertake in the national interest, to develop and support light water reactor technologies

WATER BOILER REACTOR

Science.gov (United States)

King, L.D.P.

1960-11-22

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-01-01

The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

Design study of ship based nuclear power reactor

International Nuclear Information System (INIS)

Su'ud, Zaki; Fitriyani, Dian

2002-01-01

Preliminary design study of ship based nuclear power reactors has been performed. In this study the results of thermohydraulics analysis is presented especially related to behaviour of ship motion in the sea. The reactors are basically lead-bismuth cooled fast power reactors using nitride fuels to enhance neutronics and safety performance. Some design modification are performed for feasibility of operation under sea wave movement. The system use loop type with relatively large coolant pipe above reactor core. The reactors does not use IHX, so that the heat from primary coolant system directly transferred to water-steam loop through steam generator. The reactors are capable to be operated in difference power level during night and noon. The reactors however can also be used totally or partially to produce clean water through desalination of sea water. Due to the influence of sea wave movement the analysis have to be performed in three dimensional analysis. The computation time for this analysis is speeded up using Parallel Virtual Machine (PVM) Based multi processor system

Cascade: a high-efficiency ICF power reactor

International Nuclear Information System (INIS)

Pitts, J.H.

1985-01-01

Cascade attains a net power-plant efficiency of 49% and its cost is competitive with high-temperature gas-cooled reactor, pressurized-water reactor, and coal-fired power plants. The Cascade reactor and blanket are made of ceramic materials and activation is 6 times less than that of the MARS Tandem Mirror Reactor operating at comparable power. Hands-on maintenance of the heat exchangers is possible one day after shutdown. Essentially all tritium is recovered in the vacuum system, with the remainder recovered from the helium power conversion loop. Tritium leakage external to the vacuum system and power conversion loop is only 0.03 Ci/d

Reactors of different types in the world nuclear power

International Nuclear Information System (INIS)

Simonov, K.V.

1991-01-01

The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

The heavy water accountancy for research reactors in JAERI

International Nuclear Information System (INIS)

Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

1998-11-01

The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

Guideline for examination concerning the evaluation of safety in light water power reactor installations

International Nuclear Information System (INIS)

1978-01-01

This guideline was drawn up as the guide for examination when the safety evaluation of nuclear reactors is carried out at the time of approving the installation of light water power reactors. Accordingly in case of the examination of safety, it must be confirmed that the contents of application are in conformity with this guideline. If they are in conformity, it is judged that the safety evaluation of the policy in the basic design of a reactor facility is adequate, and also that the evaluation concerning the separation from the public in surroundings is adequate as the condition of location of the reactor facility. This guideline is concerned with light water power reactors now in use, but the basic concept may be the reference for the examination of the other types of reactors. If such a case occurs that the safety evaluation does not conform to this guideline, it is not excluded when the appropriate reason is clarified. The purpose of safety evaluation, the scope to be evaluated, the selection of the events to be evaluated, the criteria for judgement, the matters taken into consideration at the time of analysis, the concrete events of abnormal transient change and accident in operation, and the concrete events of serious accident and hypothetic accident are stipulated. The explanation and two appendices are attached. (Kako, I.)

Occupational radiation exposure at light water cooled power reactors. Annual report, 1977

International Nuclear Information System (INIS)

Peck, L.J.

1979-04-01

This report presents an updated compilation of occupational radiation exposures at commercial light water cooled nuclear power reactors (LWRs) for the years 1969 through 1977. The information contained in this document was derived from reports submitted to the United States Nuclear Regulatory Commission in accordance with requirements of individual plant Technical Specifications, and in accordance with Part 20.407 of Title 10, Chapter 1, Code of Federal Regulations (10 CFR Part 20.407). An additional 4 LWRs completed a full calendar year of commercial operation for the first time in 1977. This report now encompasses data from 57 commercially operating U.S. nuclear power plants. The number of personnel monitored at LWRs increased approximately 10% in 1977, and the average collective dose to personnel (man-rems per reactor-year) increased 14% over the 1976 average. The average number of personnel receiving measurable exposure per reactor increased 11%, and the average exposure per individual in 1977 was 0.8 rem per person

To the analysis of reactor noise in boiling water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1972-01-01

The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

Management of radioactive wastes at power reactor sites in India

International Nuclear Information System (INIS)

Amalraj, R.V.; Balu, K.

Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

Natural uranium fueled light water moderated breeding hybrid power reactors: a feasibility study

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

1978-06-01

The first part of the study consists of a thorough investigation of the properties of subcritical thermal lattices for hybrid reactor applications. Light water is found to be the best moderator for (fuel-self-sufficient) FSS hybrid reactors for power generation. Several lattice geometries and compositions of particular promise for LWHRs are identified. Using one of these lattices, fueled with natural uranium, the performance of several concepts of LWHR blankets is investigated, and optimal blanket designs are identified. The effect of blanket coverage efficiency and the feasibility of separating the functions of tritium breeding and of power generation to different blankets are investigated. Optimal iron-water shields for LWHRs are also determined. The performance of generic types of LWHRs is evaluated. The evolution of the blanket properties with burnup is evaluated and fuel management schemes are briefly examined. The feasibility of using the lithium system of the blanket to control the blanket power amplitude and shape is also investigated. A parametric study of the energy balance of LWHR power plants is carried out, and performance parameters expected from LWHRs are estimated. Discussions are given of special features of LWHRs and their fuel cycle

Operational margin monitoring system for boiling water reactor power plants

International Nuclear Information System (INIS)

Fukutomi, S.; Takigawa, Y.

1992-01-01

This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

Performance of self-powered neutron detectors in pressurized water reactors

International Nuclear Information System (INIS)

Warren, H.D.; Bozarch, D.P.

1977-01-01

A typical Babcock and Wilcox pressurized water reactor (PWR) contains 364 rhodium self-powered neutron detectors (SPNDs) and 52 background detectors. The detectors are inserted into the reactor core in 52 dry, multidetector assemblies. Each assembly contains seven SPNDs and one background detector. By mid-1977, eight B and W PWRs, each fitted with SPNDs, were in operation. Many of the SPNDs have operated successfully for more than four years. This paper describes the operational performance of the SPNDs and special tests conducted to improve that performance. Topics included are (1) insulation performance versus neutron dose to the SPND, (2) background signals in the leadwire region of the SPND, and (3) depletion of the SPND emitter versus absorbed neutron dose

Inquiry into the radiological consequences of power uprates at light-water reactors worldwide

International Nuclear Information System (INIS)

Bilic Zabric, Tea; Tomic, Bojan; Lundgren, Klas; Sjoeberg, Mats

2007-05-01

In Sweden, most of the nuclear power plants are planning power uprates within the next few years. The Dept. of Occupational and Medical Exposures at the Swedish Radiation Protection Agency, SSI, has initiated a research project to investigate the radiological implications of power uprates on light-water reactors throughout the world. The project was divided into three tasks: 1. A compilation of power uprates of light-water reactors worldwide. The compilation contains a technical description in brief of how the power uprates were carried out. 2. An analysis of the radiological consequences at four selected Nuclear Power Plants, which was the main objective of the inquiry. Affects on the radiological and chemical situation due to the changed situation were discussed. 3. Review of technical and organisational factors to be considered in uprate projects to keep exposures ALARA. The project was carried out, starting with the collecting of information on the implemented and planned uprates on reactors internationally. The information was catalogued in accordance with criteria focusing on radiological impact. A detailed analysis followed of four plants selected for uprates chosen according to established criteria, in line with the project requirements. The selected plants were Olkiluoto 1 and 2, Cofrentes, Asco and Tihange. The plants were selected with design and operation conditions close to the Swedish plants. All information was compiled to identify good and bad practices that are impacting on the occupational exposure. Important factors were discussed concerning BWRs and PWRs which affect radiation levels and occupational exposures in general, and especially at power uprates. Conclusions related to each task are in detail presented in a particular chapter of the report. Taking into account the whole project and its main objective the following conclusions are considered to be emphasized: Optimisation of the work processes to limit the duration of the time spent in

Inquiry into the radiological consequences of power uprates at light-water reactors worldwide

Energy Technology Data Exchange (ETDEWEB)

Bilic Zabric, Tea; Tomic, Bojan; Lundgren, Klas; Sjoeberg, Mats

2007-05-15

In Sweden, most of the nuclear power plants are planning power uprates within the next few years. The Dept. of Occupational and Medical Exposures at the Swedish Radiation Protection Agency, SSI, has initiated a research project to investigate the radiological implications of power uprates on light-water reactors throughout the world. The project was divided into three tasks: 1. A compilation of power uprates of light-water reactors worldwide. The compilation contains a technical description in brief of how the power uprates were carried out. 2. An analysis of the radiological consequences at four selected Nuclear Power Plants, which was the main objective of the inquiry. Affects on the radiological and chemical situation due to the changed situation were discussed. 3. Review of technical and organisational factors to be considered in uprate projects to keep exposures ALARA. The project was carried out, starting with the collecting of information on the implemented and planned uprates on reactors internationally. The information was catalogued in accordance with criteria focusing on radiological impact. A detailed analysis followed of four plants selected for uprates chosen according to established criteria, in line with the project requirements. The selected plants were Olkiluoto 1 and 2, Cofrentes, Asco and Tihange. The plants were selected with design and operation conditions close to the Swedish plants. All information was compiled to identify good and bad practices that are impacting on the occupational exposure. Important factors were discussed concerning BWRs and PWRs which affect radiation levels and occupational exposures in general, and especially at power uprates. Conclusions related to each task are in detail presented in a particular chapter of the report. Taking into account the whole project and its main objective the following conclusions are considered to be emphasized: Optimisation of the work processes to limit the duration of the time spent in

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

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

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

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

New lineup of light water reactors

International Nuclear Information System (INIS)

Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

2007-01-01

Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

Improving activity transport models for water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Burrill, K.A

2001-08-01

Eight current models for describing radioactivity transport and radiation field growth around water-cooled nuclear power reactors have been reviewed and assessed. A frequent failing of the models is the arbitrary nature of the determination of the important processes. Nearly all modelers agree that the kinetics of deposition and release of both dissolved and particulate material must be described. Plant data must be used to guide the selection and development of suitable improved models, with a minimum of empirically-based rate constraints being used. Limiting case modelling based on experimental data is suggested as a way to simplify current models and remove their subjectivity. Improved models must consider the recent change to 'coordinated water chemistry' that appears to produce normal solubility behaviour for dissolved iron throughout the fuel cycle in PWRs, but retrograde solubility remains for dissolved nickel. Profiles are suggested for dissolved iron and nickel concentrations around the heat transport system in CANDU reactors, which operate nominally at constant chemistry, i.e., pH{sub T} constant with time, and which use carbon steel isothermal piping. These diagrams are modified for a CANDU reactor with stainless steel piping, in order to show the changes expected. The significance of these profiles for transport in PWRs is discussed for further model improvement. (author)

Power level effects on thorium-based fuels in pressure-tube heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

2016-06-15

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

Power level effects on thorium-based fuels in pressure-tube heavy water reactors

International Nuclear Information System (INIS)

Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

2016-01-01

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

Status of the advanced boiling water reactor and simplified boiling water reactor

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

Generic environmental impact statement on handling and storage of spent light water power reactor fuel. Appendices

International Nuclear Information System (INIS)

1978-03-01

Detailed appendices are included with the following titles: light water reactor fuel cycle, present practice, model 1000MW(e) coal-fired power plant, increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data (1976-2000), characteristics of nuclear fuel, and ''away-from-reactor'' storage concept

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

International Nuclear Information System (INIS)

Pathania, R.S.; Tatone, O.S.

1979-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

SBWR: A simplified boiling water reactor

International Nuclear Information System (INIS)

Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

1987-01-01

An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

Reactor water chemistry control

International Nuclear Information System (INIS)

Kundu, A.K.

2010-01-01

Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

Introduction to nuclear power reactors and their health physics systems

International Nuclear Information System (INIS)

Brtis, J.S.

1982-01-01

This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

International Nuclear Information System (INIS)

Boldyrev, V.M.; Mikhaj, V.I.

1985-01-01

Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

Development of an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated reactors

International Nuclear Information System (INIS)

Babaev, N.S.

1981-06-01

The results of work carried out under IAEA Contract No. 2336/RB are described (subject: an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated (VVER) reactors). The basic principles of an accounting system for this type of nuclear power plant are outlined. The general structure and individual units of the information computer program used to achieve automated accounting are described and instructions are given on the use of the program. A detailed example of its application (on a simulated nuclear power plant) is examined

Steam-generator tube failures: world experience in water-cooled nuclear power reactors in 1974

International Nuclear Information System (INIS)

Hare, M.G.

1976-01-01

Steam-generator tube failures were reported at 25 of 59 water-cooled nuclear power reactors surveyed in 1974, compared to 11 of 49 in 1973. A summary is presented of these failures, most of which, where the cause is known, were the result of corrosion. Water chemistry control, inspection and repair procedures, and failure rates are discussed

Investigation of the resonant power oscillation in the Halden Boiling Water Reactor by autoregressive modeling

International Nuclear Information System (INIS)

Oguma, Ritsuo

1980-01-01

In the HBWR (Halden Boiling Water Reactor), there exists a resonant power oscillation with period about 0.04 Hz at power levels higher than about 9.5 MWt. While the resonant oscillation in not so large as to affect the normal reactor operation, it is significant, from the viewpoint of reactor diagnosis, to grasp its characteristics and find the cause. Noise analysis based on the autoregressive (AR) modeling technique has been made to reveal the driving source for this oscillation which led to the suggestion that it is attributed to the dynamic interference of heat exchange process between two parallel-connected steam transformers against the reactor. The present study demonstrates that the method used here is highly effective for tracing back to a noise source inducing the variation of quantities in a system, and also applicable to problems of reactor noise analysis and diagnosis. (author)

Cost analysis of light water reactor power plants

International Nuclear Information System (INIS)

Mooz, W.E.

1978-06-01

A statistical analysis is presented of the capital costs of light water reactor (LWR) electrical power plants. The objective is twofold: to determine what factors are statistically related to capital costs and to produce a methodology for estimating these costs. The analysis in the study is based on the time and cost data that are available on U.S. nuclear power plants. Out of a total of about 60 operating plants, useful capital-cost data were available on only 39 plants. In addition, construction-time data were available on about 65 plants, and data on completed construction permit applications were available for about 132 plants. The cost data were first systematically adjusted to constant dollars. Then multivariate regression analyses were performed by using independent variables consisting of various physical and locational characteristics of the plants. The dependent variables analyzed were the time required to obtain a construction permit, the construction time, and the capital cost

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that the DT device. The efficienct of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed. (aurthor)

Advanced light water reactor program at ABB-Combustion Engineering Nuclear Power

International Nuclear Information System (INIS)

Cahn, H.

1990-01-01

To meet the needs of Electric Utilities ordering nuclear power plants in the 1990s, ABB-Combustion Engineering is developing two designs which will meet EPRI consensus requirements and new licensing issues. The System 80 Plus design is an evolutionary pressurized water reactor plant modelled after the successful System 80 design in operation in Palo Verde and under construction in Korea. System Plus is currently under review by the US Nuclear Regulatory Commission with final design approval expected in 1991 and design certification in 1992. The Safe Integral Reactor (SIR) plant is a smaller facility with passive safety features and modular construction intended for design certification in the late 1990s. (author)

Core design concepts for high performance light water reactors

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.

2007-01-01

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

Remerschen nuclear power station with BBR pressurized water reactor

International Nuclear Information System (INIS)

Hoffmann, J.P.

1975-01-01

On the basis of many decades of successful cooperation in the electricity supply sector with the German RWE utility, the Grand Duchy of Luxemburg and RWE jointly founded Societe Luxembourgeoise d'Energie Nucleaire S.A. (SENU) in 1974 in which each of the partners holds a fifty percent interest. SENU is responsible for planning, building and operating this nuclear power station. Following an international invitation for bids on the delivery and turnkey construction of a nuclear power station, the consortium of the German companies of Brown, Boveri and Cie. AG (BBC), Babcock - Brown Boveri Reaktor GmbH (BBR) and Hochtief AG (HT) received a letter of intent for the purchase of a 1,300 MW nuclear power station equipped with a pressurized water reactor. The 1,300 MW station of Remerschen will be largely identical with the Muelheim-Kaerlich plant under construction by the same consortium near Coblence on the River Rhine since early 1975. According to present scheduling, the Remerschen nuclear power station could start operation in 1981. (orig.) [de

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

Development Project of Supercritical-water Cooled Power Reactor

International Nuclear Information System (INIS)

Kataoka, K.; Shiga, S.; Moriya, K.; Oka, Y.; Yoshida, S.; Takahashi, H.

2002-01-01

A Supercritical-water Cooled Power Reactor (SCPR) development project (Feb. 2001- Mar. 2005) is being performed by a joint team consisting of Japanese universities and nuclear venders with a national fund. The main objective of this project is to provide technical information essential to demonstration of SCPR technologies through concentrating three sub-themes: 'plant conceptual design', 'thermohydraulics', and 'material and water chemistry'. The target of the 'plant conceptual design sub-theme' is simplify the whole plant systems compared with the conventional LWRs while achieving high thermal efficiency of more than 40 % without sacrificing the level of safety. Under the 'thermohydraulics sub-theme', heat transfer characteristics of supercritical-water as a coolant of the SCPR are examined experimentally and analytically focusing on 'heat transfer deterioration'. The experiments are being performed using fron-22 for water at a fossil boiler test facility. The experimental results are being incorporated in LWR analytical tools together with an extended steam/R22 table. Under the 'material and water chemistry sub-theme', material candidates for fuel claddings and internals of the SCPR are being screened mainly through mechanical tests, corrosion tests, and simulated irradiation tests under the SCPR condition considering water chemistry. In particular, stress corrosion cracking sensitivity is being investigated as well as uniform corrosion and swelling characteristics. Influences of water chemistry on the corrosion product characteristics are also being examined to find preferable water condition as well as to develop rational water chemistry controlling methods. (authors)

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.; Allen, E.J.

1977-01-01

A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

Pressurised water reactor operation

International Nuclear Information System (INIS)

Birnie, S.; Lamonby, J.K.

1987-01-01

The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

Super critical water reactors

International Nuclear Information System (INIS)

Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

2005-01-01

Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

Water simulation of sodium reactors

International Nuclear Information System (INIS)

Grewal, S.S.; Gluekler, E.L.

1981-01-01

The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

Simulation of decreasing reactor power level with BWR simulator

International Nuclear Information System (INIS)

Suwoto; Zuhair; Rivai, Abu Khalid

2002-01-01

Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

International conference on opportunities and challenges for water cooled reactors in the 21. century. PowerPoint presentations

International Nuclear Information System (INIS)

2009-01-01

Water Cooled Reactors have been the keystone of the nuclear industry in the 20th Century. As we move into the 21st Century and face new challenges such as the threat of climate change or the large growth in world energy demand, nuclear energy has been singled out as one of the sources that could substantially and sustainably contribute to power the world. As the nuclear community worldwide looks into the future with the development of advanced and innovative reactor designs and fuel cycles, it becomes important to explore the role Water Cooled Reactors (WCRs) will play in this future. To support the future role of WCRs, substantial design and development programmes are underway in a number of Member States to incorporate additional technology improvements into advanced nuclear power plants (NPPs) designs. One of the key features of advanced nuclear reactor designs is their improved safety due to a reduction in the probability and consequences of accidents and to an increase in the operator time allowed to better assess and properly react to abnormal events. A systematic approach and the experience of many years of successful operation have allowed designers to focus their design efforts and develop safer, more efficient and more reliable designs, and to optimize plant availability and cost through improved maintenance programs and simpler operation and inspection practices. Because many of these advanced WCR designs will be built in countries with no previous nuclear experience, it is also important to establish a forum to facilitate the exchange of information on the infrastructure and technical issues associated with the sustainable deployment of advanced nuclear reactors and its application for the optimization of maintenance of operating nuclear power plants. This international conference seeks to be all-inclusive, bringing together the policy, economic and technical decision-makers and the stakeholders in the nuclear industry such as operators, suppliers

Siting of light-water reactor power plants in the Federal Republic of Germany

International Nuclear Information System (INIS)

Kohler, H.A.G.

1975-01-01

The nuclear power plant site requirements formulated for environment protection in Germany allow nuclear power plants to be built at any site provided these requirements are duly taken into account in preparing and monitoring the site and in the design of the proposed power plant. After a brief discussion of light water reactor power plant sites, prevailing practice in site planning, site selection criteria, licensing procedure and used criteria, rules and guidelines, this paper reports on some considerations taken into account by the expert advisers and by the licensing authorities and future site planning. (orig.) [de

A Robust Multivariable Feedforward/Feedback Controller Design for Integrated Power Control of Boiling Water Reactor Power Plants

International Nuclear Information System (INIS)

Shyu, S.-S.; Edwards, Robert M.

2002-01-01

In this paper, a methodology for synthesizing a robust multivariable feedforward/feedback control (FF/FBC) strategy is proposed for an integrated control of turbine power, throttle pressure, and reactor water level in a nuclear power plant. In the proposed method, the FBC is synthesized by the robust control approach. The feedforward control, which is generated via nonlinear programming, is added to the robust FBC system to further improve the control performance. The plant uncertainties, including unmodeled dynamics, linearization, and model reduction, are characterized and estimated. The comparisons of simulation responses based on a nonlinear reactor model demonstrate the achievement of the proposed controller with specified performance and endurance under uncertainty. It is also important to note that all input variables are manipulated in an orchestrated manner in response to a single output's setpoint change

Power oscillation and stability in water cooled reactors

International Nuclear Information System (INIS)

Por, G.; Kis, G.

1998-01-01

Periodic oscillation in measured temperature fluctuation was observed near to surface of a heated rod in certain heat transfer range. The frequency of the peak found in power spectral density of temperature fluctuation and period estimated from the cross correlation function for two axially placed thermocouples change linearly with linear energy (or surface heat) production. It was concluded that a resonance of such surface (inlet) temperature oscillation with the pole of the reactor transfer function can be responsible for power oscillation in BWR and PWR, thus instability is not solely due to reactor transfer function. (author)

Normal and compact spent fuel storage in light water reactor power plants

International Nuclear Information System (INIS)

Kuenel, R.R.

1978-01-01

The compact storage of light water reactor spent fuel is a safe, cheap and reliable contribution towards overcoming the momentarily existing shortage in spent fuel reprocessing. The technical concept is described and physical behaviour discussed. The introduction of compact storage racks in nuclear power plants increases the capacity from 100 to about 240 %. The increase in decay heat is not more than about 14%, the increase in activity inventory and hazard potential does not exceed 20%. In most cases the existing power plant equipment fulfils the new requirements. (author)

Effect of water impurities on stress corrosion cracking in a boiling water reactor

International Nuclear Information System (INIS)

Ljungbery, L.G.; Cubicciotti, D

1985-01-01

A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

Safety problems of nuclear power plants with channel-type graphite boiling water reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

1977-01-01

Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

French experience in operating pressurized water reactor power stations. Ten years' operation of the Ardennes power station

International Nuclear Information System (INIS)

Teste du Bailler, A.; Vedrinne, J.F.

1978-01-01

In the paper the experience gained over ten years' operation of the Ardennes (Chooz) nuclear power station is summarized from the point of view of monitoring and control equipment. The reactor was the first pressurized water reactor to be installed in France; it is operated jointly by France and Belgium. The equipment, which in many cases consists of prototypes, was developed for industrial use and with the experience that has now been gained it is possible to evaluate its qualities and defects, the constraints which it imposes and the action that has to be taken in the future. (author)

Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1978-02-01

A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

Methods and technologies for cost reduction in the design of water cooled reactor power plants

International Nuclear Information System (INIS)

1991-05-01

The Specialists Meeting was organized in the framework of the IAEA International Working Group on Advanced Technologies for Water-Cooled Reactors. Its purpose was to provide an international forum for review and discussion on recent results in research and development on different methods and technologies of current and advanced water-cooled reactor power plants, which can lead to reduced investment and operation, maintenance and fuel-cycle costs of the plants. 27 specialists representing 10 countries and the IAEA took part in the meeting. 10 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth

2002-06-01

The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

Status of control assembly materials in Indian water reactors

International Nuclear Information System (INIS)

Date, V.G.; Kulkarni, P.G.

2000-01-01

India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-01-01

The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

How power is generated in a nuclear reactor

International Nuclear Information System (INIS)

Swaminathan, V.

1978-01-01

Power generation by nuclear fission as a result of chain reaction caused by neutrons interacting with fissile material such as 235 U, 233 U and 239 Pu is explained. Electric power production by reactor is schematically illustrated. Materials used in thermal reactor and breeder reactor are compared. Fuel reprocessing and disposal of radioactive waste coming from reprocessing plant is briefly described. Nuclear activities in India are reviewed. Four heavy water plants and two power reactors are under construction and will be operative in the near future. Two power reactors are already in operation. Nuclear Fuel Complex at Hyderabad supplies fuel element to the reactors. Fuel reprocessing and waste management facility has been set up at Tarapur. Bhabha Atomic Research Centre at Bombay and Reactor Research Centre at Kalpakkam near Madras are engaged in applied and basic research in nuclear science and engineering. (B.G.W.)

Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-06-01

Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO₂) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Risk contribution from low power and shutdown of a pressurized water reactor

International Nuclear Information System (INIS)

Chu, T.L.; Pratt, W.T.

1997-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. A phased approach was used in Level 1. In Phase 1 the concept of plant operational states (POSs) was developed to provide a better representation of the plant as it transitions from power to non power operation. This included a coarse screening analysis of all POSs to identify vulnerable plant configurations, to characterize (on a high, medium, or low basis) potential frequencies of core damage accidents, and to provide a foundation for a detailed Phase 2 analysis. In Phase 2, selected POSs from both Grand Gulf and Surry were chosen for detailed analysis. For Grand Gulf, POS 5 (approximately Cold Shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage was selected. For Surry, three POSs representing the time the plant spends in mid loop operation were chosen for analysis. Level 1 and Level 2/3 results from the Surry analyses are presented

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

Safety of light water reactors. Risks of nuclear technology

International Nuclear Information System (INIS)

Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

2012-01-01

The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

Power reactor noise

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions, thus preventing further complications by alerting opeators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Dr. Thie hopes to further, through detailed explanations and over 70 illustrations, the acceptance of the use of noise analysis by the nuclear utility industry. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussions of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

The Swedish Zero Power Reactor R0

Energy Technology Data Exchange (ETDEWEB)

Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

1961-05-15

The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

The Swedish Zero Power Reactor R0

International Nuclear Information System (INIS)

Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

1961-05-01

The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of ± 0. 1 mm

Auxiliary equipment for cooling water in a reactor

International Nuclear Information System (INIS)

Konno, Yasuhiro; Sakairi, Toshiaki.

1975-01-01

Object: To effectively make use of pressure energy of reactor water, which has heretofore been discarded, to enable supply of emergency power supply of high reliability and to prevent spreading of environmental contamination. Structure: Sea water pumped by a sea water supply pump is fed to a heat exchanger. Reactor water carried through piping on the side to be cooled is removed in heat by the heat exchanger to be cooled and returned, and then again returned to the reactor. On the other hand, sea water heated by the heat exchanger is fed to a water wheel to drive the water wheel, after which it is discharged into a discharging path. A generator may be directly connected to the water wheel to use the electricity generated by the generator as the emergency power source. (Kamimura, M.)

Economic simplified boiling water reactor (ESBWR) response to an extended station blackout/ loss of all AC power

International Nuclear Information System (INIS)

Barrett, A.J.; Marquino, W.

2013-01-01

U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackout for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

Energy Technology Data Exchange (ETDEWEB)

Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

2017-02-15

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

A novel nuclear combined power and cooling system integrating high temperature gas-cooled reactor with ammonia–water cycle

International Nuclear Information System (INIS)

Luo, Chending; Zhao, Fuqiang; Zhang, Na

2014-01-01

Highlights: • We propose a novel nuclear ammonia–water power and cooling cogeneration system. • The high temperature reactor is inherently safe, with exhaust heat fully recovered. • The thermal performances are improved compared with nuclear combined cycle. • The base case attains an energy efficiency of 69.9% and exergy efficiency of 72.5%. • Energy conservation and emission reduction are achieved in this cogeneration way. - Abstract: A nuclear ammonia–water power and refrigeration cogeneration system (NAPR) has been proposed and analyzed in this paper. It consists of a closed high temperature gas-cooled reactor (HTGR) topping Brayton cycle and a modified ammonia water power/refrigeration combined bottoming cycle (APR). The HTGR is an inherently safe reactor, and thus could be stable, flexible and suitable for various energy supply situation, and its exhaust heat is fully recovered by the mixture of ammonia and water in the bottoming cycle. To reduce exergy losses and enhance outputs, the ammonia concentrations of the bottoming cycle working fluid are optimized in both power and refrigeration processes. With the HTGR of 200 MW thermal capacity and 900 °C/70 bar reactor-core-outlet helium, the system achieves 88.8 MW net electrical output and 9.27 MW refrigeration capacity, and also attains an energy efficiency of 69.9% and exergy efficiency of 72.5%, which are higher by 5.3%-points and 2.6%-points as compared with the nuclear combined cycle (NCC, like a conventional gas/steam power-only combined cycle while the topping cycle is a closed HTGR Brayton cycle) with the same nuclear energy input. Compared with conventional separate power and refrigeration generation systems, the fossil fuel saving (based on CH 4 ) and CO 2 emission reduction of base-case NAPR could reach ∼9.66 × 10 4 t/y and ∼26.6 × 10 4 t/y, respectively. The system integration accomplishes the safe and high-efficiency utilization of nuclear energy by power and refrigeration

Water and Regolith Shielding for Surface Reactor Missions

Science.gov (United States)

Poston, David I.; Ade, Brian J.; Sadasivan, Pratap; Leichliter, Katrina J.; Dixon, David D.

2006-01-01

This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

Water and Regolith Shielding for Surface Reactor Missions

International Nuclear Information System (INIS)

Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

2006-01-01

This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density

A review of the UKAEA interest in heavy water reactors

International Nuclear Information System (INIS)

Symes, R.J.

1983-01-01

The chapter commences with a brief account of the history of heavy water production and then begins the story of the British use of this moderator in power reactors. This is equated with the introduction and development of the tube reactor as a distinct and important form of reactor construction in contrast with the perhaps better known vessel design that has tended to dominate reactor engineering to date. The account thus includes a succession of reactor designs including the gas and steam cooled heavy water systems in addition to the steam-generating heavy water reactor. The SGHWR was demonstrated by the construction of a substantial prototype, which continues in operation as a flexible and reliable electricity-generating plant. It was also, for a time, identified as the system to be used for Britain's third reactor programme. Today the successful Canadian CANDU power reactors represent the only penetration of heavy water reactor technology into large scale electricity generation. The range of research and experimental reactors using heavy water in their cores is reviewed. (author)

Channel type reactors with supercritical water coolant. Russian experience

International Nuclear Information System (INIS)

Kuznetsov, Y.N.; Gabaraev, B.A.

2003-01-01

Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

Evolution of on-power fuelling machines on Canadian natural uranium power reactors

International Nuclear Information System (INIS)

Isaac, P.

1984-10-01

The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

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

International Nuclear Information System (INIS)

Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

2008-01-01

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 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

Fuel cladding interaction with water coolant in power reactors

International Nuclear Information System (INIS)

1985-11-01

Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

Estimation, comparison, and evaluation of advanced fission power reactor generation costs

International Nuclear Information System (INIS)

Waddell, J.D.

1977-01-01

The study compares the high-temperature gas-cooled reactor (HTGR), the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the light water breeder reactor (LWBR), and the heavy water reactor (HWR) with proposed light water reactors (LWR) and liquid-metal fast breeder reactors (LMFBR). The relative electrical generation costs, including the effects of the introduction of advanced reactor fuel cycles into the U.S. nuclear power economy, were projected through the year 2030. The study utilized the NEEDS computer code which is a simulation of the U.S. nuclear power economy. The future potential electrical generation costs and cumulative consumption of uranium ore were developed using characterizations of the advanced systems. The reactor-fuel cycle characterizations were developed from literature reviews and personal discussions with the proponents of the various systems. The study developed a ranking of the concepts based on generation costs and uranium consumption

Development of light water reactors and subjects for hereafter

International Nuclear Information System (INIS)

Murao, Yoshio

1995-01-01

As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

International Nuclear Information System (INIS)

Tatone, O.S.; Meindl, P.; Taylor, G.F.

1986-06-01

A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Galperin, A; Radkowski, A [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

1996-12-01

The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors).

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

An evaluation of light water breeder reactor system (LWBR) as an alternative for nuclear power generation in Brazil

International Nuclear Information System (INIS)

Sauer, I.L.

1981-01-01

The LWBR system as an alternative for nuclear power generation in Brazil, was technically and economically evaluated. The LWBR system has been characterized comparatively with the Pressurized Water Reactors through technological and investment cost analysis and through the analysis of the processes and unit costs of the fuel cycle stages. The characteristics of the LWBR system in comparison to the PWR system, with respect to utilization and cumulative consumption of uranium and thorium resources, fuel cycle processes and associated costs have been determined for possible alternatives of nuclear power participation in the Brazilian hidro-thermal electricity generating system. The analysis concluded that the LWBR system does not represent an attractive alternative for nuclear power generation in Brazil and even has no potential to compete with conventional Pressurized Water Reactors. (Author) [pt

Graphite-water steam-generating reactor in the USSR

Energy Technology Data Exchange (ETDEWEB)

Dollezhal, N A [AN SSSR, Moscow

1981-10-01

One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

Response of pressurized water reactor (PWR) to network power generation demands

International Nuclear Information System (INIS)

Schreiner, L.A.

1991-01-01

The flexibility of the PWR type reactor in terms of response to the variations of the network power demands, is demonstrated. The factors that affect the transitory flexibility and some design prospects that allow the reactor fits the requirements of the network power demands, are also discussed. (M.J.A.)

Calibration of RB reactor power; Kalibrisanje snage reaktora RB

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Markovic, H; Ninkovic, M; Strugar, P; Dimitrijevic, Z; Takac, S; Stefanovic, D; Kocic, A; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1976-09-15

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8{radical}2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation.

Thorium in heavy water reactors

International Nuclear Information System (INIS)

Andersson, G.

1984-12-01

Advanced heavy water reactors can provide energy on a global scale beyond the foreseeable future. Their economic and safety features are promising: 1. The theoretical feasibility of the Self Sufficient Equilibrium Thorium (SSET) concept is confirmed by new calculations. Calculations show that the adjuster rod geometry used in natural uranium CANDU reactors is adequate also for SSET if the absorption in the rods is graded. 2. New fuel bundle designs can permit substantially higher power output from a CANDU reactor. The capital cost for fuel, heavy water and mechanical equipment can thereby be greatly reduced. Progress is possible with the traditional fuel material oxide, but the use of thorium metal gives much larger effects. 3. A promising long range possibility is to use pressure tanks instead of pressure tubes. Heat removal from the core is facilitated. Negative temperature and void coefficients provide inherent safety features. Refuelling under power is no longer needed if control by moderator displacement is used. Reduced quality demand on the fuel permits lower fuel costs. The neutron economy is improved by the absence of pressure and clandria tubes and also by the use of radial and axial blankets. A modular seed blanket design can reduce the Pa losses. The experience from construction of tank designs is good e.g. AAgesta, Attucha. It is now also possible to utilize technology from LWR reactors and the implementation of advanced heavy water reactors would thus be easier than HTR or LMFBR systems. (Author)

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

Electrochemistry of Water-Cooled Nuclear Reactors

International Nuclear Information System (INIS)

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

2006-01-01

This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

Taiwan Power Company's power distribution analysis and fuel thermal margin verification methods for pressurized water reactors

International Nuclear Information System (INIS)

Huang, P.H.

1995-01-01

Taiwan Power Company's (TPC's) power distribution analysis and fuel thermal margin verification methods for pressurized water reactors (PWRs) are examined. The TPC and the Institute of Nuclear Energy Research started a joint 5-yr project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, these methods were developed to allow TPC to independently perform verifications of the local power density and departure from nucleate boiling design bases, which are required by the reload safety evaluation for the Maanshan PWR plant. The computer codes utilized were extensively validated for the intended applications. Sample calculations were performed for up to six reload cycles of the Maanshan plant, and the results were found to be quite consistent with the vendor's calculational results

Pressurized water reactors: the EPR project

International Nuclear Information System (INIS)

Py, J.P.; Yvon, M.

2007-01-01

EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

International Nuclear Information System (INIS)

Whitehead, D.W.; Hake, T.M.

1990-01-01

Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

Safety of next generation power reactors

International Nuclear Information System (INIS)

Anon.

1988-01-01

This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1999-11-01

In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

Recent computer applications in boiling water reactor power plants

International Nuclear Information System (INIS)

Hiraga, Shoji; Joge, Toshio; Kiyokawa, Kazuhiro; Kato, Kanji; Nigawara, Seiitsu

1976-01-01

Process computers in boiling water reactor power plants have won the position of important equipments for the calculation of the core and plant performances and for data logging. Their application technique is growing larger and larger every year. Here, two systems are introduced; plant diagnostic system and computerized control panel. The plant diagnostic system consists of the part processing the signals from a plant, the operation part mainly composed of a computer to diagnose the operating conditions of each system component using input signal, and the result display (CRT or typewriter). The concept on the indications on control panels in nuclear power plants is changing from ''Plant parameters and to be indicated on panel meters as much as possible'' to ''Only the data required for operation are to be indicated.'' Thus the computerized control panel is attracting attention, in which the process computer for processing the operating information and CRT display are introduced. The experimental model of that panel comprises and operator's console and a chief watchmen's console. Its functions are dialogic data access and the automatic selection of preferential information. (Wakatsuki, Y.)

Water cooled type nuclear power plant

International Nuclear Information System (INIS)

Arai, Shigeki.

1981-01-01

Purpose: To construct high efficiency a PWR type nuclear power plant with a simple structure by preparing high temperature and pressure water by a PWR type nuclear reactor and a pressurizer, converting the high temperature and high pressure water into steam with a pressure reducing valve and introducing the steam into a turbine, thereby generating electricity. Constitution: A pressurizer is connected downstream of a PWR type nuclear reactor, thereby maintaining the reactor at high pressure. A pressure-reducing valve is provided downstream of the pressurizer, the high temperature and pressure water is reduced in pressure, thereby producing steam. The steam is fed to a turbine, and electric power is generated by a generator connected to the turbine. The steam exhausted from the turbine is condensed by a condenser into water, and the water is returned through a feedwater heater to the reactor. Since the high temperature and pressure water in thus reduced in pressure thereby evaporating it, the steam can be more efficiently produced than by a steam generator. (Sekiya, K.)

Method of operating a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Lysell, G.

1975-01-01

When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

Reactor water quality degradation suppressing method upon reactor start up

International Nuclear Information System (INIS)

Maeda, Katsuharu.

1993-01-01

Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

Fuzzy algorithm for an automatic reactor power control in a PWR

International Nuclear Information System (INIS)

Hah, Yung Joon; Song, In Ho; Yu, Sung Sik; Choi, Jung In; Lee, Byong Whi

1994-01-01

A fuzzy algorithm is presented for automatic reactor power control in a pressurized water reactor. Automatic power shape control is complicated by the use of control rods because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability for the load - follow operation including frequency control. In an attempt to achieve automatic power shape control without any design modification of the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multi - input multi - output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to the Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of the pressurized water reactor during the load - follow operation

The control of emissions from nuclear power reactors in Canada

International Nuclear Information System (INIS)

Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

1988-01-01

Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

Supercritical-pressure, once-through cycle light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi

2001-01-01

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

Power Excursion Accident Analysis of Research Water Reactor

International Nuclear Information System (INIS)

Khaled, S.M.; Doaa, G.M.

2009-01-01

A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

Safety considerations concerning light water reactors in Sweden

International Nuclear Information System (INIS)

Nilsson, T.

1977-01-01

In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

International Nuclear Information System (INIS)

Brand, P.; Wiechers, W.K.

1974-01-01

The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

Several perspectives on water-chemical cycles for nuclear power stations equipped with type VVER and RBMK reactors

International Nuclear Information System (INIS)

Mamet, A.P.; Mamet, V.A.; Pashevich, V.I.; Nazarenko, P.N.

1982-01-01

Water-chemical cycles for loops I and II of VVER reactors are discussed. These cycles are mixed ammonia-sodium with a variable concentration of boric acid and ammonia hydrazine with a pH factor of 9.1 +/- 0.1. New water-chemical cycles are considered for use in both existing and new nuclear power plants. Application of these new water-chemical cycles showed produce a significant improvement in operating conditions of nuclear power plants. Upon accumulation of sufficient operating experience with these cycles, it should be possible to raise the issue of revising applicable standard documentation

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Light water reactor safety research project

International Nuclear Information System (INIS)

Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

1980-07-01

The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

Dual pressurized light water reactor producing 2000 M We

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

Functional systems of a pressurized water reactor

International Nuclear Information System (INIS)

Heinzel, V.

1982-01-01

The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

Tokamak experimental power reactor

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

1978-01-01

A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

The IAEA power reactor information system - PRIS

International Nuclear Information System (INIS)

Laue, H.J.; Qureshi, A.; Skjoeldebrand, R.; White, D.

1983-01-01

The IAEA Power Reactor Information System, PRIS, is based on a collection of basic design data and operating experience data which the IAEA started in 1970. PRIS is used for annual publications on 'Power Reactors in Member States', 'Operating Experience with Nuclear Power Stations in Member States', which gives annual operating information for individual plants, and a 'Performance Analysis Report' summarizing each year's and earlier experience. Since 1973 information has been collected in a systematic manner on significant plant outages (= more than 10 full power hours). There is now information on more than 10,000 outages in the system which permits some conclusions to be drawn both in regard to individual plants and to categories of plants on the significance of different outage reasons and different types of equipment failures. PRIS has not been intended to be a component reliability information system as an international data collection must stop short of the level of detail which would be needed for that purpose. The objectives of PRIS have been to provide a factual background for assumptions on parameters which are essential for economic evaluations and for systems operation planning (load factor and availability). The outage information does, however, lend itself to conclusions about generic problems in different categories of plants and it can be used by an individual operator to find other plants where information about particular problems can be obtained. It would also now be possible to use PRIS for setting availability goals based on experience and not only on theoretical design considerations. The paper demonstrates the conclusions which can be drawn from 662 reactor years of operation of light and heavy water pressurized reactors and 390 reactor years of boiling water reactors and, in particular, the role that the main heat removal system and its components have played in the equipment failure category

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Radiation streaming in power reactors. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

1979-02-01

Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

Control of operational transients in power reactors - Methodology

International Nuclear Information System (INIS)

Vukovic, D.

1983-01-01

By introducing the nuclear power stations in the electric power system, questions of their possibilities to satisfy system's demand arise. Control of operational transients (temperature and Xe 135 ) in power reactors by determining the optimal control rod strategy is given. Ti optimize the Xe 135 transients, the Pantryagin theorem of optimal processes is applied. For solving three dimensional, two-group diffusion equations the heterogeneous Feinberg-Galanin method with axial flux harmonics is adopted. An application of this formalism to three-dimensional, finite cylindrical pressurised water reactor radially reflected is presented. (author)

New advanced small and medium nuclear power reactors: possible nuclear power plants for Australia

International Nuclear Information System (INIS)

Dussol, R.J.

2003-01-01

In recent years interest has increased in small and medium sized nuclear power reactors for generating electricity and process heat. This interest has been driven by a desire to reduce capital costs, construction times and interest during construction, service remote sites and ease integration into small grids. The IAEA has recommended that the term 'small' be applied to reactors with a net electrical output less than 300 MWe and the term 'medium' to 300-700 MWe. A large amount of experience has been gained over 50 years in the design, construction and operation of small and medium nuclear power reactors. Historically, 100% of commercial reactors were in these categories in 1951-1960, reducing to 21% in 1991-2000. The technologies involved include pressurised water reactors, boiling water reactors, high temperature gas-cooled reactors, liquid metal reactors and molten salt reactors. Details will be provided of two of the most promising new designs, the South African Pebble Bed Modular Reactor (PBMR) of about 110 MWe, and the IRIS (International Reactor Innovative and Secure) reactor of about 335 MWe. Their construction costs are estimated to be about US$l,000/kWe with a generating cost for the PBMR of about US1.6c/kWh. These costs are lower than estimated for the latest designs of large reactors such as the European Pressurised Reactor (EPR) designed for 1,600 MWe for use in Europe in the next decade. It is concluded that a small or medium nuclear power reactor system built in modules to follow an increasing demand could be attractive for generating low cost electricity in many Australian states and reduce problems arising from air pollution and greenhouse gas emissions from burning fossil fuels

Power control of water reactors using nitrogen 16 activity measurements

International Nuclear Information System (INIS)

Gariod, R.; Merchie, F.; O'byrne, G.

1964-01-01

At the Grenoble Nuclear Research Centre, the open-core swimming pool reactors Melusine (2 MW) and Siloe (15 MW) are controlled at a constant overall power using nitrogen-16 channels. The conventional linear control channels react instantaneously to the rapid power fluctuations, this being necessary for the safety of the reactors, but their power indications are erroneous since they are affected by local deformations of the thermal flux caused by the compensation movements of the control rods. The nitrogen-16 channels on the other hand give an indication of the overall power proportional to the mean fission flux and independent of the rod movements, but their response time is 15 seconds, A constant overall power control is thus possible by a slow correction of the reference signal given by the automatic control governed by thu linear channels by means of a correction term given by the 'N-16' channels: This is done automatically in Melusine and manually in Siloe. (authors) [fr

Thermal and stability considerations for a supercritical water-cooled fast reactor during power-raising phase of plant startup

International Nuclear Information System (INIS)

Cai, Jiejin; Ishiwatari, Yuki; Oka, Yoshiaki; Ikejiri, Satoshi

2009-01-01

This paper describes thermal analyses and linear stability analyses of the Supercritical Water-cooled Fast Reactor with 'two-path' flow scheme during the power-raising phase of plant startup. For thermal consideration, the same criterion of the maximum cladding surface temperature (MCST) as applied to the normal operating condition is used. For thermal-hydraulic stability consideration, the decay ratio of 0.5 is applied, which is taken from BWRs. Firstly, we calculated the flow rate distribution among the parallel flow paths from the reactor vessel inlet nozzles to the mixing plenum below the core using a system analysis code. The parallel flow paths consist of the seed fuel assemblies cooled by downward flow, the blanket fuel assemblies cooled by downward flow and the downcomer. Then, the MCSTs are estimated for various reactor powers and feedwater flow rates with system analyses. The decay ratios are estimated with linear stability analyses. The available range of the reactor power and feedwater flow rate to satisfy the thermal and stability criteria is obtained. (author)

New approach for control rod position indication system for light water power reactor

International Nuclear Information System (INIS)

Bahuguna, Sushil; Dhage, Sangeeta; Nawaj, S.; Salek, C.; Lahiri, S.K.; Marathe, P.P.; Mukhopadhyay, S.; Taly, Y.K.

2015-01-01

Control rod position indication system is an important system in a nuclear power plant to monitor and display control rod position in all regimes of reactor operation. A new approach to design a control rod position indication system for sensing absolute position of control rod in Light Water Power Reactor has been undertaken. The proposed system employs an inductive type, hybrid measurement strategy providing both analog position as well as digital zone indication with built-in temperature compensation. The new design approach meets single failure criterion through redundancy in design without sacrificing measurement resolution. It also provides diversity in measurement technique by indirect position sensing based on analysis of drive coil current signature. Prototype development and qualification at room temperature of the control rod position indication system (CRPIS) has been demonstrated. The article presents the design philosophy of control rod position indication system, the new measurement strategy for sensing absolute position of control rod, position estimation algorithm for both direct and indirect sensing and a brief account associated processing electronics. (author)

The next generation of power reactors - safety characteristics

International Nuclear Information System (INIS)

Modro, S.M.

1995-01-01

The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1981-01-01

The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Some particular aspects of control in nuclear power reactors

International Nuclear Information System (INIS)

Vathaire, F. de; Vernier, Ph.; Pascouet, A.

1964-01-01

This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [fr

Auxiliary water supply device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

Development of small and medium reactors for power and heat production

International Nuclear Information System (INIS)

Becka, J.

1978-01-01

Data are given on the current state of development of small and medium-power reactors designed mainly for electric power production in small power grids, for heat production for small- and medium-power desalination plants with possible electric power generation, for process steam production and heat development for district heating systems, again combined with electric power generation, and for propelling big and fast passenger ships. A diagram is shown of the primary system of an integrated PWR derived from the Otto Hahn reactor. The family is listed of the standard sizes of the integral INTERATOM company pressurized water reactors. Also listed are the specifications and design of CAS 2CG and AS 3G type reactors used mainly for long-distance heating systems. (J.B.)

Nuclear reactor capable of electric power generation during in-service inspection

International Nuclear Information System (INIS)

Nakamura, Shinsuke; Nogami, Hitoshi.

1992-01-01

The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

Reactor power control device

International Nuclear Information System (INIS)

Doi, Kazuyori.

1981-01-01

Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

Simulation of fuel rods vibration in power reactors by vibration of tape coated with cadmium

International Nuclear Information System (INIS)

Holland, L.

1982-01-01

The circulation of cooling water in light water power reactor makes a vibration in internal components. The monitoring of those vibrations is necessary aiming to the safety use of reactors. Aiming at study those vibrations a neutron absorber, type vibratory tape was introduced in the core of a research reactor type Pulstar, operating at 80 W of power. The induced power variations were measured with an ionization chamber put besides the reactor core. The detector signal was recorded and analysed in a PDP-11 computer. The analysis of the results show that the power density of the detector signal, and thus, the power reactor, increase in the O-25 Hz range with an increase in the pulse height vibration. (E.G.) [pt

Safety aspects of pressurised water reactors

International Nuclear Information System (INIS)

1985-01-01

This submission to the Health and Safety Executive has been prepared by the Institution of Professional Civil Servants (IPCS) as a contribution to the debate on safety aspects associated with Pressurized Water Reactors (PWRs). Although supporting an energy policy which includes the development of nuclear power, assurances are sought on a number of safety issues if it is decided that this should be generated by a PWR-type reactor. These issues are listed. In particular the following are mentioned: the wider publication of design information, the use of elastic-plastic fracture mechanics as the basis for determining pressure vessel integrity, the failure rate of steam generating units, water coolant quality control, greater investigation of two-phase flow accident conditions, the components of the reactor cooling system and training of reactor personnel in the understanding of LOCA effects. (U.K.)

ULTRA SCWR+: Practical advanced water reactor concepts

International Nuclear Information System (INIS)

Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

2008-01-01

Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

Fuel Summary Report: Shippingport Light Water Breeder Reactor

International Nuclear Information System (INIS)

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

Penn State advanced light water reactor concept

International Nuclear Information System (INIS)

Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

1987-01-01

The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

Directory of Open Access Journals (Sweden)

Li Gang

2016-01-01

Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Science.gov (United States)

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

A comparative analysis of the domestic and foreign licensing processes for power and non-power reactors

International Nuclear Information System (INIS)

Joe, J. C.; Youn, Y. K.; Kim, W. S.; Kim, H. J.

2003-01-01

The System-integrated Modular Advanced Reactor (SMART), a small to medium sized integral type Pressurized Water Reactor (PWR) has been developed in Korea. Now, SMART-P, a 1/5 scaled-down of the SMART, is being developed for the purpose of demonstrating the safety and performance of SMART design. The SMART-P is a first-of-a-kind reactor which is utilized for the research and development of a power reactor. Since the licensing process of such a reactor is not clearly specified in the current Atomic Energy Act, a comparative survey and analysis of domestic and foreign licensing processes for power and non-power reactors has been carried out to develop the rationale and technical basis for establishing the licensing process of such a reactor. The domestic and foreign licensing processes of power and non-power reactors have been surveyed and compared, including those of the U.S.A., Japan, France, U.K., Canada, and IAEA. The general trends in nuclear reactor classification, licensing procedures, regulatory technical requirements, and other licensing requirements and regulations have been investigated. The results of this study will be used as the rationale and technical basis for establishing the licensing process of reactors at development stage such as SMART-P

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

Reliability of reactor plant water cleanup pumps

International Nuclear Information System (INIS)

Pearson, J.L.

1979-01-01

Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

Power ramp testing method for PWR fuel rod at research reactor

International Nuclear Information System (INIS)

Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

2003-01-01

A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

Advanced water-cooled reactor technologies. Rationale, state of progress and outlook

International Nuclear Information System (INIS)

Anon.

1989-01-01

Eighty per cent of the world's power reactors are water cooled and moderated. Many improvements in their design and operation have been implemented since the first such reactor started commercial operation in 1957. This report addresses the safety, environmental and economic rationales for further improvements, as well as their relevance to currently operating water reactors

Nonlinear dynamics of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

1983-01-01

Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

Indian experience with radionuclide transport, deposition and decontamination in water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Narasimhan, S.V.; Das, P.C.; Lawrence, D.A.; Mathur, P.K.; Venkateswarlu, K.S.

1983-01-01

The present generation of water-cooled nuclear reactors uses construction materials chosen with utmost care so that minimum corrosion occurs during the life of the reactor. As interaction between the primary coolant and the construction materials is unavoidable, the coolant is chemically treated to achieve maximum compatibility. First measurements of the chemical and radiochemical composition of the crud present on the in-core and out-of-core primary heat transport system surfaces of a pressurized heavy-water-moderated and cooled reactor (PHWR) are given; then experience in India in the development of a low temperature, one-stage decontaminating formulation for chemical decontamination of the radioactive deposits formed on stainless steel surfaces under BWR conditions is discussed. The effect of the magnitude of the transients in parameters such as reactor power, system temperature, dissolved oxygen content in the coolant, etc. on the nature and migration behaviour of primary heat transport system crud in a PHWR is described. Contributions to radioactive sources and insoluble crud from different primary heat transport system materials are identified and correlated with reactor operations in a PHWR. Man-rem problems faced by nuclear reactors, especially during off-line maintenance, stress the need for reducing the deposited radioactive sources from system surfaces which would otherwise be accessible. Laboratory and on-site experimentation was carried out to effect chemical decontamination on the radioactive deposits formed on the stainless steel surfaces under BWR conditions. Both the reducing and oxidizing formulations were subsequently used in a small-scale, in-plant trial in the clean-up system of a BWR. More than 85% of the deposited 60 Co activity was found to have been removed by the oxidizing formulation. Efforts to develop a decontaminating mixture containing a reducing agent with the help of a circulating loop are in progress in the laboratory. (author)

Proven power reactor systems - novel features and developments in operation performance, safety and reliability

International Nuclear Information System (INIS)

Bugl, J.

1975-01-01

As the development of nuclear reactors for the generation of electric power started after the end of the Second World War, the prospective use of diverse materials as fuel, moderator and coolant resulted in a wide diversity of design possibilities. Of the 10 nuclear reactor types which were being considered most seriously in those days, only a few have achieved acceptance. This development is best illustrated by listing the nuclear power plants in service, under construction and on order at present, separately by reactor types (table). In the lead at present and for some years to come are the thermal reactors and especially the light water reactors (LWR). In the LWR group the lead is held by the pressurised water reactor (PWR) which accounts for 44% of the installed capacity of all the nuclear power plants in service at present. In the early 1980s this share will increase to 58%, whereas the share of the boiling water reactor (BWR) will increase to only 28% from 23% at present. (orig./TK) [de

An optimized power conversion system concept of the integral, inherently-safe light water reactor

International Nuclear Information System (INIS)

Memmott, Matthew J.; Wilding, Paul R.; Petrovic, Bojan

2017-01-01

Highlights: • Three power conversion systems (PCS) for the I"2S-LWR are presented. • An optimization analyses was performed to evaluate these PCS alternatives. • The ideal PCS consists of 5 turbines, and obtains an overall efficiency of 35.7%. - Abstract: The integral, inherently safe light water reactor (I"2S-LWR) has been developed to significantly enhance passive safety capabilities while maintaining cost competitiveness relative to the current light water reactor (LWR) fleet. The compact heat exchangers of the I"2S-LWR preclude boiling of the secondary fluid, which decreases the probability of heat exchanger failure, but this requires the addition of a flash drum, which negatively affects the overall plant thermodynamic efficiency. A state of the art Rankine cycle is proposed for the I"2S-LWR to increase the thermodynamic efficiency by utilizing a flash drum with optimized operational parameters. In presenting this option for power conversion in the I"2S-LWR power plant, the key metric used in rating the performance is the overall net thermodynamic efficiency of the cycle. In evaluating the flash-Rankine cycle, three basic industrial concepts are evaluated, one without an intermediate pressure turbine, one with an intermediate turbine and one reheat stream, and one with an intermediate turbine and two reheat streams. For each configuration, a single-path multi-variable optimization is undertaken to maximize the thermal efficiency. The third configuration with an intermediate turbine and 2 reheat streams is the most effective concept, with an optimized efficiency of 35.7%.

Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

International Nuclear Information System (INIS)

Shwageraus, E.; Fridman, E.

2008-01-01

In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO 2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO 2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

Reactor power monitoring device

International Nuclear Information System (INIS)

Dogen, Ayumi; Ozawa, Michihiro.

1983-01-01

Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

Limit regulation system for pressurized water nuclear reactors

International Nuclear Information System (INIS)

Aleite, W.; Bock, H.W.

1976-01-01

Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

Power maximization of a spheric reflected reactor with optimized fuel distribution

International Nuclear Information System (INIS)

Reade, Joamar Rodrigues Vincent

1979-01-01

The maximum power of a spheric reflected reactor was determined using the theory of optimal control. The control variable employed was the fuel distribution, in accordance to constraints on the power density and on the concentration fuel. It was considered a thermal reactor with a fixed radius. The reactor was fuelled with U-235 and moderated with light water. The nuclear reactor was described by a diffusion theory model. The analytical solution was obtained for both two and four groups of energy and a FORTRAN program was developed to obtain the numerical results. (author)

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

Preliminary nuclear power reactor technology qualitative assessment for Malaysia

International Nuclear Information System (INIS)

Shamsul Amri Sulaiman

2011-01-01

Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

Water vapor as a perspective coolant for fast reactors

International Nuclear Information System (INIS)

Kalafati, D.D.; Petrov, S.I.

1978-01-01

Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Fuzzy controller for real time supervision of nuclear power reactor

International Nuclear Information System (INIS)

Bala Subramanian, R.

2012-01-01

Generally nuclear energy provides about 60% of the whole electricity production. A modulation of the nuclear power plants must be able to respond to the demand on the network. The pressurized water nuclear reactor has to yield correctly a load set point. Fundamentally, two parameters are concerned in leading this task to a successful conclusion: the power axial-offset and the control rods position. The focus of this study is the automation of the control of the power axial-offset by adding soluble boron and by minimizing the volume flows through the water pump. It is also important to take into consideration the liquid waste volume. Water or boron is injected into the reactor primary circuit. At the present time this task is still performed manually by an operator, for all previous attempts to automate it failed. That device, sketchily described in the paper, gave rise to the development of a real-time fuzzy controller for the power axial-offset and the control rods insertion in a pressurized water reactor (PWR). The fuzzy controller, which is the main subject of the paper, expresses more naturally the human expertise, thus avoiding the previous issue of empirical tunings. It was implemented in simulation using Matlab-Simulink on a Sun workstation. Two realistic tests discussed show that the fuzzy controller runs as efficiently as an expert operator does

Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

Reactor power control method upon accidents of electrical power system

International Nuclear Information System (INIS)

Hirose, Masao.

1983-01-01

Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

Power reactors in member states

International Nuclear Information System (INIS)

1975-01-01

This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

Directory of Open Access Journals (Sweden)

Gholam Reza Ansarifar

2015-12-01

Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

International Nuclear Information System (INIS)

Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

2001-01-01

The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

Study of advanced fission power reactor development for the United States. Volume I

International Nuclear Information System (INIS)

1976-01-01

This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

Steady state thermal hydraulic analysis of a boiling water reactor core, for various power distributions, using computer code THABNA

International Nuclear Information System (INIS)

Venkat Raj, V.; Saha, D.

1976-01-01

The core of a boiling water reactor may see different power distributions during its operational life. How some of the typical power distributions affect some of the thermal hydraulic parameters such as pressure drop minimum critical heat flux ratio, void distribution etc. has been studied using computer code THABNA. The effect of an increase in the leakage flow has also been analysed. (author)

Reactor power reduction system and method

International Nuclear Information System (INIS)

Bruno, S.J.; Dunn, S.A.; Raber, M.

1978-01-01

A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

Modular Lead-Bismuth Fast Reactors in Nuclear Power

Directory of Open Access Journals (Sweden)

Vladimir Petrochenko

2012-09-01

Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

Local Fission Gas Release and Swelling in Water Reactor Fuel during Slow Power Transients

DEFF Research Database (Denmark)

Mogensen, Mogens Bjerg; Walker, C.T.; Ray, I.L.F.

1985-01-01

Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two...... factors that determine the level of fission gas release during a power bump. Release begins when gas bubbles on grain boundaries start o interlink. This occurred at r/r0 ~ 0.75. Release tunnels were fully developed at r/r0 ~ 0.55 with the result that gas release was 60–70% at this position....

Design of virtual SCADA simulation system for pressurized water reactor

International Nuclear Information System (INIS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-01-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

Design of virtual SCADA simulation system for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

2016-02-08

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

Hydrogen in water-cooled nuclear power reactors

International Nuclear Information System (INIS)

1992-01-01

The Commission of the European Community (CEC) and the International Atomic Energy Agency (IAEA) decided in 1989 to update the state of the art concerning hydrogen in water cooled nuclear power reactors by commissioning a report which would review, all the available information to-date and make recommendations for the future. This joint report was prepared by committees formed by the IAEA and by the CEC. The aim of this report is to review the current understanding on the areas in which the research on hydrogen in LWR is conventionally presented, taking into account the results of the latest reported research developments. The main reactions through which hydrogen is produced are assessed together with their timings. An estimation of the amount of hydrogen produced by each reaction is given, in order to reckon their relative contribution to the hazard. An overview is then given of the state of knowledge of the most important phenomena taking place during its transport from the place of production and the phenomena which control the hydrogen combustion and the consequences of combustion under various conditions. Specific research work is recommended in each sector of the presented phenomena. The last topics reviewed in this report are the hydrogen detection and the prevent/mitigation of pressure and temperature loads on containment structures and structures and safety related equipment caused by hydrogen combustion

Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

2008-07-01

In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

A dual pressurized water reactor producing 2000 MWe

International Nuclear Information System (INIS)

Kang, K. M.; Suh, K. Y.

2010-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

A stability identification system for boiling water nuclear reactors

International Nuclear Information System (INIS)

Belblidia, L.A.; Chevrier, A.

1994-01-01

Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Malkoske, G.R.; Norton, J.L.; Slack, J.

2002-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Progress Report for Year 1, Quarter 2 (January - March 2002)

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-03-01

The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

Fundamentals of boiling water reactor (BWR)

International Nuclear Information System (INIS)

Bozzola, S.

1982-01-01

These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

An estimation of reactor thermal power uncertainty using UFM-based feedwater flow rate in nuclear power plants

International Nuclear Information System (INIS)

Byung Ryul Jung; Ho Cheol Jang; Byung Jin Lee; Se Jin Baik; Woo Hyun Jang

2005-01-01

Most of Pressurized Water Reactors (PWRs) utilize the venturi meters (VMs) to measure the feedwater (FW) flow rate to the steam generator in the calorimetric measurement, which is used in the reactor thermal power (RTP) estimation. However, measurement drifts have been experienced due to some anomalies on the venturi meter (generally called the venturi meter fouling). The VM's fouling tends to increase the measured pressure drop across the meter, which results in indication of increased feedwater flow rate. Finally, the reactor thermal power is overestimated and the actual reactor power is to be reduced to remain within the regulatory limits. To overcome this VM's fouling problem, the Ultrasonic Flow Meter (UFM) has recently been gaining attention in the measurement of the feedwater flow rate. This paper presents the applicability of a UFM based feedwater flow rate in the estimation of reactor thermal power uncertainty. The FW and RTP uncertainties are compared in terms of sensitivities between the VM- and UFM-based feedwater flow rates. Data from typical Optimized Power Reactor 1000 (OPR1000) plants are used to estimate the uncertainty. (authors)

Ultrasonic level and temperature sensor for power reactor applications

International Nuclear Information System (INIS)

Dress, W.B.; Miller, G.N.

1983-01-01

An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel

Application of fuzzy logic control system for reactor feed-water control

International Nuclear Information System (INIS)

Iijima, T.; Nakajima, Y.

1994-01-01

The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Hursin, M.; Perret, G. [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland)

2012-07-01

The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

Monitoring device for the power distribution within a nuclear reactor core

International Nuclear Information System (INIS)

Tanzawa, Tomio; Kumanomido, Hironori; Toyoshi, Isamu.

1986-01-01

Purpose: To provide a monitoring device for the power distribution in the reactor core that calculates the power distribution based on the measurement by instruments disposed within the reactor core of BWR type reactors. Constitution: The power distribution monitoring device in a reactor core comprises a signal correcting device, a signal normalizing device and a power distribution calculating device, in which the power distribution calculating device is constituted with an average power calculating device for four fuel assemblies and an average power calculating device for fuel assemblies. Gamma-ray signals corrected by the signal correcting device and signals from neutron detectors are inputted to the signal normalizing device, both of which are calibrated to determine the axial gamma-ray signal distribution in the central water gap region with the four fuel assemblies being as the unit. The average power from the four fuel assemblies are inputted to the fuel assembly average power calculating device to allocate to each of the fuel assembly average power thereby attaining the purpose. Further, thermal restriction values are calculated thereby enabling to secure the fuel integrity. (Kamimura, M.)

Reactor power distribution monitor

International Nuclear Information System (INIS)

Hoizumi, Atsushi.

1986-01-01

Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

Economic competitiveness requirements for evolutionary water cooled reactors

International Nuclear Information System (INIS)

Hudson, C.R.; Bertel, E.; Paik, K.H.; Roh, J.H.; Tort, V.

1999-01-01

This paper analyses the necessary economic conditions for evolutionary water cooled reactors to be competitive. Utilising recent national cost data for fossil-fired base load plants expected to be commissioned by 2005 -2010, target costs for nuclear power plants are discussed. Factors that could contribute to the achievement of those targets by evolutionary water cooled reactors are addressed. The feed-back from experience acquired in implementing nuclear programmes is illustrated by some examples from France and the Republic of Korea. The paper discusses the impacts on nuclear power competitiveness of globalisation and deregulation of the electricity market and privatisation of the electricity sector. In addition, issues related to external cost internalisation are considered. (author)

Experiences on reduction of reactor water silica and fresh resin leaching organics for Kuo-Sheng Nuclear Power Plant

International Nuclear Information System (INIS)

Wen, T-J.; Wang, C-H.

2010-01-01

The silica level in reactor water of Kuo-Sheng nuclear power plants has been slowly increased from 200 ppb to the high level above 500 ppb in recent years. The results obtained from steam/liquid mass balance calculation indicated that an increase of reactor water silica was mainly caused by continuing equilibrium leakage from deep bed condensate demineralizers, where the ion exchange zone was periodically disturbed by resin backwashing - scrubbing operation. The fastest and the most effective way to reduce the silica inventory in reactor system is to operate by continuously precoating of two sets of the reactor water clean up filter demineralizers to a lower effluent silica end point, and perhaps as frequently as three or four days. Leaching organic contaminants into feed water from the ion exchange resin becomes a key greater problem of current concern for the stable water quality promotion of condensate demineralizer. The presence of those impurities have practically been difficult to analyze by simple quality testing of the resin, and may result in as much as a hundred fold increase in chloride and sulfate in reactor water. As resin displacement with high leachable TOC, a repeated continuous soaking and effectively rinsing is required so that steady state TOC content less than 150 ppb should be achieved in an acceptably short period of time before put in-service. It is clear that cation resin containing high leachables generates high level of sulfates and sometimes also gives unexpected level of chlorides. The current TOC limits in condensate demineralizer effluent with 0.1 ppb become a significant experience to maintain reactor water soluble impurity in low levels. New resin should be subjected to TOC quality control testing prior to acceptance especially when first placed into service. TOC and organic chloride leachables for as-received virgin cation resin that are to be used in condensate polisher should be limited to be less than 100 mg-TOC and 0.5 mg-Cl per

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

International Nuclear Information System (INIS)

1995-08-01

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

Data list of nuclear power plants of pressurized-water reactor type in Japan

International Nuclear Information System (INIS)

Izumi, Fumio; Harayama, Yasuo

1981-08-01

This report has collected and compiled the data concerning performances, equipments and installations for nuclear power plants of the pressurized-water reactor type in Japan. The data used in the report are based on informations that were collected before December in 1980. The report is edited by modifing changes of the data appeared after publication of 1979 edition (JAERI-M 8947), and extending the data-package to cover new plants proposed thereafter. All data have been processed and tabulated with a computer program FREP, which has been developed as an exclusive use of data processing. (author)

Utility requirements for advanced light water reactors

International Nuclear Information System (INIS)

Machiels, A.; Gray, S.; Mulford, T.; Rodwell, E.

1996-01-01

The nuclear energy industry is actively engaged in developing advanced light water reactor (ALWR) designs for the next century. The new designs take advantage of the thousands of reactor-years of experience that have been accumulated by operating over 400 plants worldwide. The EPRI effort began in the early 1980's, when a survey of utility executives was conducted to determine their prerequisites for ordering nuclear power plants. The results were clear: new plants had to be simpler and safer, and have greater design margins, i.e., be more forgiving. The utility executives also supported making improvements to the established light water reactor technology, rather than trying to develop new reactor concepts. Finally, they wanted the option to build mid-size plants (∼600 MWe) in addition to full-size plants of more than 1200 MWe. 4 refs

TRIGA research reactors with higher power density

International Nuclear Information System (INIS)

Whittemore, W.L.

1994-01-01

The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (≅48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D 2 O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 10 14 n/cm 2 ·s and fast fluxes (>0.1 MeV) of 2 x 10 14 n/cm 2 ·s. The core centerline thermal neutron flux in the D 2 O reflector is about 2 x 10 14 n/cm 2 ·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D 2 O reflector of 3 x 10 14 n/cm 2 ·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

Use of reactor plants of enhanced safety for sea water desalination, industrial and district heating

International Nuclear Information System (INIS)

Panov, Yu.; Polunichev, V.; Zverev, K.

1997-01-01

Russian designers have developed and can deliver nuclear complexes to provide sea water desalination, industrial and district heating. This paper provides an overview of these designs utilizing the ABV, KLT-40 and ATETS-80 reactor plants of enhanced safety. The most advanced nuclear powered water desalination project is the APVS-80. This design consists of a special ship equipped with the distillation desalination plant powered at a level of 160 MW(th) utilizing the type KLT-40 reactor plant. More than 20 years of experience with water desalination and reactor plants has been achieved in Aktau and Russian nuclear ships without radioactive contamination of desalinated water. Design is also proceeding on a two structure complex consisting of a floating nuclear power station and a reverse osmosis desalination plant. This new technology for sea water desalination provides the opportunity to considerably reduce the specific consumption of power for the desalination of sea water. The ABV reactor is utilized in the ''Volnolom'' type floating nuclear power stations. This design also features a desalinator ship which provides sea water desalination by the reverse osmosis process. The ATETS-80 is a nuclear two-reactor cogeneration complex which incorporates the integral vessel-type PWR which can be used in the production of electricity, steam, hot and desalinated water. (author). 9 figs

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Tritium issues in commercial pressurized water reactors

International Nuclear Information System (INIS)

Jones, G.

2008-01-01

Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

Water quality maintaining device of power plant

International Nuclear Information System (INIS)

Kobayashi, Minoru; Inami, Ichiro.

1994-01-01

The device of the present invention reduces the amount of leaching materials of ion exchange resins from a water processing system of a BWR tyep plant, improves the water quality of reactor water to maintain the water at high purity. That is, steams used for power generation are condensated in a condensate system. A condensate filter and a condensate desalter for cleaning the condensates are disposed. A resin storage hopper is disposed for supplying the ion exchange resins to the water processing system. A device for supplying a nitrogen gas or an inert gas is disposed in the hopper. With such a constitution, the ion exchange resins in the water processing system are maintained in a nitrogen gas or inert gas atmosphere or at a low dissolved oxygen level in an operation stage in the power plant. Accordingly, degradation of the ion exchange resins in the water processing system is suppressed and the amount of the leaching material from the resins is reduced. As a result, the amount of the resins leached into the reactor is reduced, so that the reactor water quality can be maintained at high purity. (I.S.)

Development of methods for monitoring and controlling power in nuclear reactors

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole dos; Silva, Vitor Vasconcelos Araujo

2012-01-01

Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 TRIGA nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235 U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 TRIGA reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of

Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

Science.gov (United States)

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-01-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

Advances in heavy water reactors

International Nuclear Information System (INIS)

1994-03-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

International Nuclear Information System (INIS)

Kang, Kyoung Min; Noh, Sang Woo; Suh, Kune Yull

2009-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the 21 st century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

Modeling and simulation of pressurized water reactor power plant

International Nuclear Information System (INIS)

Wang, S.J.

1983-01-01

Two kinds of balance of plant (BOP) models of a pressurized water reactor (PWR) system are developed in this work - the detailed BOP model and the simple BOP model. The detailed model is used to simulate the normal operational performance of a whole BOP system. The simple model is used to combine with the NSSS model for a whole plant simulation. The trends of the steady state values of the detailed model are correct and the dynamic responses are reasonable. The simple BOP model approach starts the modelling work from the overall point of view. The response of the normalized turbine power and the feedwater inlet temperature to the steam generator of the simple model are compared with those of the detailed model. Both the steady state values and the dynamic responses are close to those of the detailed model. The simple BOP model is found adequate to represent the main performance of the BOP system. The simple balance of plant model was coupled with a NSSS model for a whole plant simulation. The NSSS model consists of the reactor core model, the steam generator model, and the coolant temperature control system. A closed loop whole plant simulation for an electric load perturbation was performed. The results are plausible. The coupling effect between the NSSS system and the BOP system was analyzed. The feedback of the BOP system has little effect on the steam generator performance, while the performance of the BOP system is strongly affected by the steam flow rate from the NSSS

Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

International Nuclear Information System (INIS)

1989-02-01

Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1987-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1988-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicate but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: Nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat. (orig.)

LOFA analyses for the water and helium cooled SEAFP reactors

International Nuclear Information System (INIS)

Sponton, L.; Sjoeberg, A.; Nordlinder, S.

2001-01-01

This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

Reactor power control system

International Nuclear Information System (INIS)

Tomisawa, Teruaki.

1981-01-01

Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

Approximation model of three-dimensional power distribution in boiling water reactor using neural networks

International Nuclear Information System (INIS)

Kobayashi, Yoko; Aiyoshi, Eitaro

2001-01-01

Fast and accurate prediction of three-dimensional (3D) power distribution is essential in a boiling water reactor (BWR). The prediction method of 3D power distribution in BWR is developed using the neural network. Application of the neural network starts with selecting the learning algorithm. In the proposed method, we use the learning algorithms based on a class of Quasi-Newton optimization techniques called Self-Scaling Variable Metric (SSVM) methods. Prediction studies were done for a core of actual BWR plant with octant symmetry. Compared to classical Quasi-Newton methods, it is shown that the SSVM method reduces the number of iterations in the learning mode. The results of prediction demonstrate that the neural network can predict 3D power distribution of BWR reasonably well. The proposed method will be very useful for BWR loading pattern optimization problems where 3D power distribution for a huge number of loading patterns (LPs) must be performed. (author)

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

International Nuclear Information System (INIS)

Noh, Hyung Gyun; Kang, Hie Chan

2016-01-01

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Kang, Hie Chan [Kunsan University, Gunsan (Korea, Republic of)

2016-05-15

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

Analysis of thermal fatigue events in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)

General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

International Nuclear Information System (INIS)

Roche, R.; Gaudez, J.C.

1964-01-01

In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

STUDY OF WATER HAMMERS IN THE FILLING OF THE SYSTEM OF PRESSURE COMPENSATION IN THE WATER-COOLED AND WATER-MODERATED POWER REACTORS

Directory of Open Access Journals (Sweden)

A. V. Korolyev

2017-01-01

Full Text Available The research presented in the article conforms to the severe accident that took place at the Three Mail Island nuclear power plant in the USA. The research is focused on improving the reliability of the pressure compensator that is an important equipment of the primary circuit. In order to simulate such a situation, the stand has been developed to simulate the design of the pressurizer of the PWR-440 reactor, in particular an elliptical shape of the upper lid which has a steam outlet pipe at the top of the construction that creates conditions for occurrence of such water hammers. For the experiments, an installation has been created that makes it possible to measure and record the water hammering that occur when the tanks are filled. Measurement of the amplitude of the water hammering was carried out by a specially developed piezoelectric sensor, and the registration – by a light-beam oscilloscope. The technique of carrying out the experiment is described and the results of an experimental study of the water hammers arising when the vessels are completely filled are presented. Quantitative data were obtained on the amplitudes of the hydraulic impacts depending on the rate of filling of the vessel and the diameter of the outlet, the maximum pressure of the hydraulic shock was 7–9 atm. Comparison of calculated and experimental data has been performed. The allowable discrepancy is explained by the calculated value of the system stiffness coefficient, which did not take into account the presence of welded seams in the tank that imparts the system with additional rigidity. The calculated relationships are obtained, that make it possible to estimate the amplitudes of the water hammers through the acceleration of the water level in front of the outlet from a vessel with an elliptical bottom. The possibility of a water hammer in the pressure compensator is demonstrated by experiment and by theoretical calculations. Based on the experimental data, a

Thermal hydraulic aspects of uncertainty in power measurement of nuclear reactors

International Nuclear Information System (INIS)

Gupta, S.K.; Kumar, Rajesh; Gaikwad, A.J.; Majumdar, P.; Agrawal, R.A.

2004-01-01

Power measurement in Nuclear Reactors is carried out through in-core and ex-core neutron monitors which are continuously calibrated against thermal power. In Indian Pressurized Heavy Water Reactors (220 MWe) the temperature difference across steam generator hot and cold legs is taken to be a measure of thermal power as the flow through the primary heat transport system is assumed to be constant through out is operation. Gross flow is not measured directly. However, the flow depends on the characteristics of the primary heat transport pumps, which are centrifugal type and are affected by the grid frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable high grid frequency. This uncertainty is in addition to instrument inaccuracy and should be accounted for in safety analysis. In some reactors thermal power is calculated from stem flow rate and pressure, here the location of steam flow measurement is important to avoid leakage related error in thermal power. Neutron absorption cross section in the power measurement instruments and the power production in the fuel varies with neutron energy levels, these aspects are also discussed in the paper. (author)

The Consortium for Advanced Simulation of Light Water Reactors

International Nuclear Information System (INIS)

Szilard, Ronaldo; Zhang, Hongbin; Kothe, Douglas; Turinsky, Paul

2011-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

Review of irradiation experiments for water reactor safety research

International Nuclear Information System (INIS)

Tobioka, Toshiaki

1977-02-01

A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

Small size modular fast reactors in large scale nuclear power

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Dragunov, U.G.; Stepanov, V.S.; Klimov, N.N.; Kopytov, I.I.; Krushelnitsky, V.N.

2005-01-01

The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

Sea water take-up facility for cooling reactor auxiliary

International Nuclear Information System (INIS)

Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

1997-01-01

The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

Program of RA reactor start-up to nominal power

International Nuclear Information System (INIS)

1959-01-01

The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is described in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members

Low power modular power generating reactors or Small Modular Reactors (SMR)

International Nuclear Information System (INIS)

Chenais, Jacques

2016-01-01

Electronuclear reactors were small reactors at the beginning, and then tend to be always bigger and more powerful, but since some recent times, several countries specialized in reactor design and fabrication (USA, Russia, China, and South Korea) have been developing Small Modular Reactors (SMR) of less than 300 MW. As France has already produced feasibility studies and is about to launch a SMR development programme, the author comments some specific aspects of this new architecture of reactors, characterises the targeted markets, gives an overview of the various more or less advanced existing concepts: a floating barge in Russia, the SMART 100 MW project in South Korea, several concepts in the USA (the mPower 125 MW, the NuScale 45 MW, the Westinghouse 225 MW, and the HI-SMUR 160 MW projects), the ACP 100 MW in China, the CAREM 27 MW in Argentina. French projects developed by the CEA, EDF, Areva and DCNS are then presented

Reactor water sampling device

International Nuclear Information System (INIS)

Sakamaki, Kazuo.

1992-01-01

The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

High temperature on-line monitoring of water chemistry and corrosion control in water cooled power reactors. Report of a co-ordinated research project 1995-1999

International Nuclear Information System (INIS)

2002-07-01

This report documents the results of the Co-ordinated Research Project (CRP) on High Temperature On-line Monitoring of Water Chemistry and Corrosion in Water Cooled Power Reactors (1995-1999). This report attempts to provide both an overview of the state of the art with regard to on-line monitoring of water chemistry and corrosion in operating reactors, and technical details of the important contributions made by programme participants to the development and qualification of new monitoring techniques. The WACOL CRP is a follow-up to the WACOLIN (Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors) CRP conducted by the IAEA from 1986 to 1991. The WACOLIN CRP, which described chemistry, corrosion and activity-transport aspects, clearly showed the influence of water chemistry on corrosion of both fuel and reactor primary-circuit components, as well as on radiation fields. It was concluded that there was a fundamental need to monitor water-chemistry parameters in real time, reliably and accurately. The objectives of the WACOL CRP were to establish recommendations for the development, qualification and plant implementation of methods and equipment for on-line monitoring of water chemistry and corrosion. Chief investigators from 18 organizations representing 15 countries provided a variety of contributions aimed at introducing proven monitoring techniques into plants on a regular basis and filling the gaps between plant operator needs and available monitoring techniques. The CRP firmly demonstrated that in situ monitoring is able to provide additional and valuable information to plant operators, e.g. ECP, high temperature pH and conductivity. Such data can be obtained promptly, i.e. in real time and with a high degree of accuracy. Reliable techniques and sensor devices are available which enable plant operators to obtain additional information on the response of structural materials in

Reactor water clean-up device

International Nuclear Information System (INIS)

Tanaka, Koji; Egashira, Yasuo; Shimada, Fumie; Igarashi, Noboru.

1983-01-01

Purpose: To save a low temperature reactor water clean-up system indispensable so far and significantly simplify the system by carrying out the reactor water clean-up solely in a high temperature reactor water clean-up system. Constitution: The reactor water clean-up device comprises a high temperature clean-up pump and a high temperature adsorption device for inorganic adsorbents. The high temperature adsorption device is filled with amphoteric ion adsorbing inorganic adsorbents, or amphoteric ion adsorbing inorganic adsorbents and anionic adsorbing inorganic adsorbents. The reactor water clean-up device introduces reactor water by the high temperature clean-up pump through a recycling system to the high temperature adsorption device for inorganic adsorbents. Since cations such as cobalt ions and anions such as chlorine ions in the reactor water are simultaneously removed in the device, a low temperature reactor water clean-up system which has been indispensable so far can be saved to realize the significant simplification for the entire system. (Seki, T.)

An aqueous lithium salt blanket option for fusion power reactors

Energy Technology Data Exchange (ETDEWEB)

Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

1989-04-01

An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Fast reactors in nuclear power

Energy Technology Data Exchange (ETDEWEB)

Kazachkovskii, O

1981-02-01

The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

The design features of integrated modular water reactor (IMR)

International Nuclear Information System (INIS)

Kanagawa, T.; Goto, M.; Usui, S.; Suzuta, T.; Serizawa, A.; Kunugi, T.; Yamauchi, T.; Itoh, G.; Matsumura, T.

2004-01-01

Small-to-medium-sized (300-600 MWe) reactors are required for the electric power market in the near future (2010-2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and small-scale design is needed. From such point of view, Integrated Modular Water Reactor (IMR), whose electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of the reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid systems, and Instrumentation and Control systems of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented. (authors)

Expert system for maintenance management of a boiling water reactor power plant

International Nuclear Information System (INIS)

Hong Shen; Liou, L.W.; Levine, S.; Ray, A.; Detamore, M.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, which must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components

Analysis of an accelerator-driven subcritical light water reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

1997-01-01

An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

2012-09-01

To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

TerraPower, Bill Gates' reactor

International Nuclear Information System (INIS)

Guidez, J.

2016-01-01

TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

Qualification of the Taiwan Power Company's pressurized water reactor physics methods using CASMO-4/SIMULATE-3