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Sample records for reactor plant prototype

  1. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  2. Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

    International Nuclear Information System (INIS)

    1984-06-01

    The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

  3. Prototype moving-ring reactor

    International Nuclear Information System (INIS)

    Smith, A.C. Jr.; Ashworth, C.P.; Abreu, K.E.

    1981-01-01

    The objective of this work was to design a prototype fusion reactor based on fusion plasmas confined as ''Compact Toruses.' Six major criteria guided the prototype design. The prototype must: (1) produce net electricity decisively (P/sub net/ >70% of P/sub gross/), with P/sub net/ approximately 100 MW(e); (2) have small physical size (low project cost) but commercial plant; (3) have all features required of commerical plants; (4) avoid unreasonable extrapolation of technology; (5) minimize nuclear issues substantially, i.e. accident and waste issues of public concern, and (6) be modular (to permit repetitive fabrication of parts) and be maintainable with low occupational radiological exposures

  4. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This report includes engineering memorandums, drawings, key feature descriptions, and other data. Some of the reports, such as manufacturability and some stress analysis, were done by consultants for Byron Jackson. Review of this report indicates that the design is feasible. The pump can be manufactured to system and specification requirements. The overall length and weight of some pieces will require special consideration, but is within the scope of equipment and technology available today. The fabricated parts are large and heavy, but can be manufactured and machined. Only the high temperature is unique to this size, since previous sodium pumps were smaller. Nondestructive tests as required by the Code are described and are feasible. The performance test of the prototype has been studied thoroughly. It is feasible for a cold water test. There are some problem areas. However, all of them can be solved. Development needs include building and testing a small scale model.

  5. Prototype vibration measurement program for reactor internals (177-fuel assembly plant). Supplement 1

    International Nuclear Information System (INIS)

    Simonis, J.C.; Post, R.C.; Thoren, D.E.

    1976-08-01

    The surveillance specimen holder tubes installed in the Babcock and Wilcox 177-fuel assembly plants have been redesigned. The structural adequacy of this design has been verified through extensive analysis. The design adequacy will be further confirmed by measuring the vibrational response of the surveillance specimen holder tube during normal and transient flow operation. This report describes the vibration measurement program that will be conducted at Toledo Edison's Davis Besse 1 site

  6. Prototype moving-ring reactor

    International Nuclear Information System (INIS)

    Smith, A.C. Jr.; Ashworth, C.P.; Abreu, K.E.

    1982-01-01

    We have completed a design of the Prototype Moving-Ring Reactor. The fusion fuel is confined in current-carrying rings of magnetically-field-reversed plasma (Compact Toroids). The plasma rings, formed by a coaxial plasma gun, undergo adiabatic magnetic compression to ignition temperature while they are being injected into the reactor's burner section. The cylindrical burner chamber is divided into three burn stations. Separator coils and a slight axial guide field gradient are used to shuttle the ignited toroids rapidly from one burn station to the next, pausing for 1/3 of the total burn time at each station. D-T- 3 He ice pellets refuel the rings at a rate which maintains constant radiated power

  7. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1996-01-01

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  8. Development of prototype reactor maintenance. (2) Application to piping support of sodium-cooled reactor prototype

    International Nuclear Information System (INIS)

    Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji; Ito, Takaya; Yamaguchi, Akira

    2017-01-01

    A maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of piping supports could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports. (author)

  9. Design of micro-reactors and solar photocatalytic prototypes

    International Nuclear Information System (INIS)

    Flores E, R.M.; Hernandez H, M.; Perusquia del Cueto, M.R.; Bonifacio M, J.; Jimenez B, J.; Ortiz O, H.B.; Castaneda J, G.; Lugo H, M.

    2007-01-01

    In the ININ is carried out research in heterogeneous photocatalysis using artificial light for to degrade organic compounds. In this context, it is sought to use the solar radiation as energy source to knock down costs. Of equal form it requires to link the basic and applied research. For it, a methodology that allows to design and to build micro-reactors and plants pilot has been developed, like previous step, to request external supports and to a future commercialization. The beginning of these works gave place to the partial construction of a prototype of photocatalytic reactor of the cylinder-parabolic composed type (CPC)

  10. Description of the prototype fast reactor at Dounreay

    International Nuclear Information System (INIS)

    Jensen, S.E.; Oelgaard, P.L.

    1995-12-01

    The Prototype Fast Reactor (PFR) at Dounreay, UK, started operation in 1975 and was closed down in 1994. The present report contains a description of the PFR nuclear power plant, based on information available in literature and on information supplied during a visit to the plant. The report covers a description of the site and plant arrangement, the buildings and structures, the reactor core and other vessel internals, the control system, the main cooling system, the decay heat removal system, the emergency core cooling system, the containment system, the steam and power conversion system, the fuel handling system, plant safety features, the control and instrumentation systems and the sodium purification systems. (au) 16 refs

  11. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  12. Operation management of the prototype heavy water reactor 'Fugen'

    International Nuclear Information System (INIS)

    Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi

    1983-01-01

    The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported. (Kako, I.)

  13. Conceptual design of reactor assembly of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

    1996-01-01

    The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

  14. An abnormal event advisory expert system prototype for reactor operators

    International Nuclear Information System (INIS)

    Hance, D.C.

    1989-01-01

    Nuclear plant operators must respond correctly during abnormal conditions in the presence of dynamic and potentially overwhelming volumes of information. For this reason, considerable effort has been directed toward the development of nuclear plant operator aids using artificial intelligence techniques. The objective of such systems is to diagnose abnormal conditions within the plant, possibly predict consequences, and advise the operators of corrective actions in a timely manner. The objective of the work is the development of a prototype expert system to diagnose abnormal events at a nuclear power plant and advise plant operators of the event and applicable procedures in an on-line mode. The major difference between this effort and previous work is the use of plant operating procedures as a knowledge source and as an integral part of the advice provided by the expert system. The acceptance by utilities of expert systems as operator aids requires that such systems be compatible with the regulatory environment and provide economic benefits. For this reason, commercially viable operator aid systems developed in the near future must complement existing plant procedures rather than reach beyond them in a revolutionary manner. A knowledge source is the resource providing facts and relationships that are coded into the expert system program. In this case, the primary source of knowledge is a set of selected abnormal operating procedures for a modern Westinghouse pressurized water reactor

  15. The stand prototype of minimum power NRE reactor

    International Nuclear Information System (INIS)

    Belogurov, A.I.; Grigorenko, L.N.; Mamontov, Yu.I.; Rachuk, V.S.; Stukalov, A.I.; Konyukhov, G.V.

    1995-01-01

    For ensuring of full-scale development of nuclear rocket engine (NRE) reactor was created stand prototype (reactor IRGIT?) The main differences of its are as follows: 1) Fasteners of technologies channels contents fuel assemblies in bottom are worked out the split. It is provides possibility a distance channels change without disassembly of reactor stand prototype from stand; 2) Cooling of the vessels, the moderator, the reflector and the barrel actuate is carried out by hydrogen; 3) The lower bottom modified for organization the hydrogen efflux in the form a reactor jet; 4) Radiation defence is introduced as part of stand prototype for ensuring of serviceability of stand accessories and tests routine service; 5) Each technology channels is provided of critical nozzle; 6) Control, regulation and defence of reactor has being carried out on stand system

  16. Pre evaluation for heat balance of prototype sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Han, Ji Woong; Kim, De Hee; Yoon, Jung; Kim, Eui Kwang; Lee, Tae Ho

    2012-01-01

    Under the long term advanced SFR R and D plan, the design of prototype reactor has been carried out toward the construction of the prototype SFR plant by 2028. The R and D efforts in fluid system design will be focused on developing a prototype design of primary heat transport system(PHTS), intermediate heat transport system (IHTS), decay heat removal system(DHRS), steam generation system(SGS), and related auxiliary system design for a prototype reactor as shown in Fig. 1. In order to make progress system design, top tier requirements for prototype reactor related to design parameters of NSSS and BOP should be decided at first. The top tier requirement includes general design basis, capacity and characteristics of reactor, various requirements related to safety, performance, securities, economics, site, and etc.. Extensive discussion has been done within Korea Atomic Energy Research Institute(KAERI) for the decision of top tier requirements of the prototype reactor. The core outlet temperature, which should be described as top tier requirements, is one of the critical parameter for system design. The higher core exit temperature could contribute to increase the plant efficiency. However, it could also contribute to decrease the design margin for structure and safety. Therefore various operating strategies based on different core outlet temperatures should be examined and evaluated. For the prototype reactor two core outlet temperatures are taken into accounted. The lower temperature is for the operation condition and the higher temperature is for the system design and licensing process of the prototype reactor. In order to evaluate the operability of prototype reactor designed based on higher temperature, the heat balance calculations have been performed at different core outlet temperature conditions. The electrical power of prototype reactor was assumed to be 100MWe and reference operating conditions were decided based on existing available data. The

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

    International Nuclear Information System (INIS)

    Simon, M.; Baust, E.; Schoening, J.

    1986-10-01

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

  18. Three-dimensional finite-element analysis of the cellular convection phenomena in the Clinch River Breeder Reactor Plant prototype pump

    International Nuclear Information System (INIS)

    Silver, A.H.; Lee, J.Y.

    1983-01-01

    Cellular convection was studied rigorously during the development of the Clinch River Breeder Reactor Plant (CRBRP) Program Pumps. This paper presents the development of a three-dimensional finite-element heat transfer model which accounts for the cellular convection phenomena. A buoyancy driven cellular convection flow pattern is introduced in the annulus region between the upper inner structure and the pump tank. Steady-state thermal data were obtained for several test conditions for argon gas pressures up to 93 psig (741 kPa) and sodium operating temperatures to 1000 0 F (811 0 K). Test temperature distributions on the pump tank and inner structure were correlated with numerical results and excellent agreement was obtained

  19. Ground testing of an SP-100 prototypic reactor

    International Nuclear Information System (INIS)

    Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

    1988-01-01

    SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

  20. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  1. Harmonizing the prototypes concerning the fast reactors of 4. generation

    International Nuclear Information System (INIS)

    Anon.

    2008-01-01

    In january 2008, an agreement was signed between the Japan Atomic Energy Agency (JAEA), the American Department of Energy (DOE) and the French Atomic Energy Commission, in order to harmonize the projects of the 3 countries for the development of prototypes of sodium-cooled fast reactors. This cooperation concerns the following issues: -) the purpose of the prototypes, -) common set of safety rules, -) technical innovations for reducing construction, operating and maintenance costs, and -) information exchange about the level of power, the type of nuclear fuels and the time schedule of these prototypes. (A.C.)

  2. The Atomics International (AI) prototype large breeder reactor (PLBR)

    International Nuclear Information System (INIS)

    McDonald, J.S.; Campise, A.V.; Brunings, J.

    1978-01-01

    The AI-PLBR breeder plant design prepared for ERDA and EPRI is of 1000 MWe size, utilizing a loop-type sodium system configuration and producing 2200 psig/850 0 F steam. A 'bullseye' core geometry type sodium system configuration and is employed with Pu0 2 -UO 2 fuel and UO 2 fertile material. The reactor outlet coolant temperature is 930 0 F. A modified 'A'-frame refueling system is employed, which is capable of handling 1/3 of a core loading in 12 days. An inducer-type mechanical pump is used in the primary circuit because of its excellent NPSH characteristics. Hockey sticks steam generators are used to produce near-fossil steam conditions at the turbine throttle. The balance of plant (BOP) design was developed by the architectural-engineering firm of Burns and Roe. It includes Allis-Chalmers - KWU 1800-rpm tandem-compound turbine, selected with high-,intermediate-, and two double-flow low-pressure cylinders equipped with two moisture separator cyclones. A design feature to enhance the licensability of the reference design is the three-level Decay Heat Removal System (DHRS), which consists of normal, backup, and diverse decay heat removal paths. The Atomics International PLBR design illustrated the technical soundness of the LMFBR system for meeting the world's long-term electrical energy supply needs. The design includes design features to assure licensability and innovative engineering features to enhance reliability, constructability, economics, and U.S. utility grid compatibility. The design concept provides a sound basis for the future detailed design of a prototype plant and subsequent development of larger LMFBR plants. (author)

  3. Safety requirements expected to the prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    2014-11-01

    In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

  4. Challenges and achievements - Prototype Fast Breeder Reactor construction

    International Nuclear Information System (INIS)

    Subramani, V.A.; Dhere, S.S.; Manoharan, V.; Subbaraman, P.

    2010-01-01

    Prototype fast breeder reactor presently under construction poses several challenges in materials, design and construction. The civil structure and equipment are of very large size and complex in nature. This paper presents the features of the design and construction of the PFBR excavation, raft, civil structure of the nuclear island connected buildings and reactor vault. This paper also brings out the details of the large size equipment of special stainless steel and handling structure for their lifting and placement inside the reactor vault. The paper is divided into three parts viz. introduction, challenges and achievements during construction of civil structures and erection of large size components. (author)

  5. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

    2002-01-01

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6 / ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  6. Some post operational adjustments to the prototype fast reactor at Dounreay

    International Nuclear Information System (INIS)

    Lunt, A.R.W.

    1979-01-01

    Prior to and during the initial operation of the Prototype Fast Reactor at Dounreay certain features have been considered to be in need of adjustment to provide better operating characteristics. This article describes the work done to support the consequential changes of operational techniques and plant design in the following areas: maintenance of dry conditions at the superheater steam inlets, the temperature control of the reactor roof, and the introduction of a system enabling the reactor to continue running after a turbine trip. (author)

  7. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  8. Development of a fresh plutonium fuel container for a prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Ohtake, T.; Takahashi, S.; Mishima, T.; Kurakami, J.; Yamamoto, Y.; Ohuchi, Y.

    1989-01-01

    Japan gives a good deal of encouragement to development of a fast breeder reactor (which is considered as the most likely candidate for nuclear power generation) to secure long-term energy source. And, following an experimental fast breeder reactor Joyo, a prototype fast breeder reactor Monju is now under vigorous construction. Related to development of the prototype fast breeder reactor, it is necessary and important to develop transport container which is used for transporting fresh fuel assemblies from Plutonium Fuel Production Facility to the Monju power plant. Therefore, the container is now being developed by Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, shipment and vibration tests, handling performance tests, shielding performance tests and prototype container tests are executed with prototype containers fabricated according to a final design, in order to experimentally confirm soundness of transport container and its contents, and propriety of design technique. This paper describes the summary of general specifications and structures of this container and the summary of preliminary safety analysis of package

  9. Comparative Studies of Core Thermal Hydraulic Design Methods for the Prototype Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Lim, Jae Yong; Kim, Sang Ji

    2013-01-01

    In this work, various core thermal-hydraulic design methods, which have arisen during the development of a prototype SFR, are compared to establish a proper design procedure. Comparative studies have been performed to determine the appropriate design method for the prototype SFR. The results show that the minimization method show a lower cladding midwall temperature than the fixed outlet temperature methods and superior thermal safety margin with the same coolant flow. The Korea Atomic energy Research Institute (KAERI) has performed a conceptual SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damages in SFR subassemblies are arisen from a creep induced failure. The creep limit is evaluated based on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, the core thermalhydraulic design method, which eventually determines the cladding temperature, is highly important to assure a safe and reliable operation of the reactor systems

  10. Irradiation of an uranium silicide prototype in RA-3 reactor

    International Nuclear Information System (INIS)

    Calabrese, R.; Estrik, G.; Notari, C.

    1996-01-01

    The factibility of irradiation of an uranium silicide (U 3 Si 2 ) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U 3 O 8 as fissible material. The enrichment of both standard and prototype is the same: 20% U 235 and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs

  11. Operability design review of prototype large breeder reactor (PLBR) designs. Final report, September 1981

    International Nuclear Information System (INIS)

    Beakes, J.H.; Ehman, J.R.; Jones, H.M.; Kinne, B.V.T.; Price, C.M.; Shores, S.P.; Welch, J.K.

    1981-09-01

    Prototype Large Breeder Reactor (PLBR) designs were reviewed by personnel with extensive power plant operations experience. Fourteen normal and off-normal events, such as startup, shutdown, refueling, reactor scram and loss of feedwater, were evaluated using an operational evaluation methodology which is designed to facilitate talk-through sessions on operational events. Human factors engineers participated in the review and assisted in developing and refining the review methodologies. Operating experience at breeder reactor facilities such as Experimental Breeder Reactor-II (EBR-II), Enrico Fermi Atomic Power Plant - Unit 1, and the Fast Flux Test Facility (FFTF) was gathered, analyzed, and used to determine whether lessons learned from operational experience had been incorporated into the PLBR designs. This eighteen month effort resulted in approximately one hundred specific recommendations for improving the operability of PLBR designs

  12. Quantity and management of spent fuel from prototype and research reactors in Germany

    International Nuclear Information System (INIS)

    Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang; Tholen, Marion

    2013-01-01

    Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on the information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)

  13. Design of micro-reactors and solar photocatalytic prototypes; Diseno de micro-reactores y prototipos fotocataliticos solares

    Energy Technology Data Exchange (ETDEWEB)

    Flores E, R.M.; Hernandez H, M.; Perusquia del Cueto, M.R.; Bonifacio M, J.; Jimenez B, J.; Ortiz O, H.B.; Castaneda J, G.; Lugo H, M. [ININ, Km. 36.5 Carr. Mexico-Toluca, 52750 La Marquesa, Ocoyoacac (Mexico)]. e-mail: rmfe@nuclear.inin.mx

    2007-07-01

    In the ININ is carried out research in heterogeneous photocatalysis using artificial light for to degrade organic compounds. In this context, it is sought to use the solar radiation as energy source to knock down costs. Of equal form it requires to link the basic and applied research. For it, a methodology that allows to design and to build micro-reactors and plants pilot has been developed, like previous step, to request external supports and to a future commercialization. The beginning of these works gave place to the partial construction of a prototype of photocatalytic reactor of the cylinder-parabolic composed type (CPC)

  14. Nuclear instrumentation systems in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Vijayakumaran, P.M.; Nagaraj, C.P.; Paramasivan-Pillai, C.; Ramakrishnan, R.; Sivaramakrishna, M.

    2004-01-01

    The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection and Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM (safety action) to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident. (authors)

  15. Operation management of the prototype heavy water reactor 'Fugen'

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

    1983-09-01

    The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported.

  16. U Plant Geographic Zone Cleanup Prototype

    International Nuclear Information System (INIS)

    Romine, L.D.; Leary, K.D.; Lackey, M.B.; Robertson, J.R.

    2006-01-01

    The U Plant geographic zone (UPZ) occupies 0.83 square kilometers on the Hanford Site Central Plateau (200 Area). It encompasses the U Plant canyon (221-U Facility), ancillary facilities that supported the canyon, soil waste sites, and underground pipelines. The UPZ cleanup initiative coordinates the cleanup of the major facilities, ancillary facilities, waste sites, and contaminated pipelines (collectively identified as 'cleanup items') within the geographic zone. The UPZ was selected as a geographic cleanup zone prototype for resolving regulatory, technical, and stakeholder issues and demonstrating cleanup methods for several reasons: most of the area is inactive, sufficient characterization information is available to support decisions, cleanup of the high-risk waste sites will help protect the groundwater, and the zone contains a representative cross-section of the types of cleanup actions that will be required in other geographic zones. The UPZ cleanup demonstrates the first of 22 integrated zone cleanup actions on the Hanford Site Central Plateau to address threats to groundwater, the environment, and human health. The UPZ contains more than 100 individual cleanup items. Cleanup actions in the zone will be undertaken using multiple regulatory processes and decision documents. Cleanup actions will include building demolition, waste site and pipeline excavation, and the construction of multiple, large engineered barriers. In some cases, different cleanup actions may be taken at item locations that are immediately adjacent to each other. The cleanup planning and field activities for each cleanup item must be undertaken in a coordinated and cohesive manner to ensure effective execution of the UPZ cleanup initiative. The UPZ zone cleanup implementation plan (ZCIP) [1] was developed to address the need for a fundamental integration tool for UPZ cleanup. As UPZ cleanup planning and implementation moves forward, the ZCIP is intended to be a living document that will

  17. The first Swedish nuclear reactor - from technical prototype to scientific instrument; Sveriges foersta kaernreaktor - fraan teknisk prototyp till vetenskapligt instrument

    Energy Technology Data Exchange (ETDEWEB)

    Fjaestad, M. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of History of Science and Technology

    2001-01-01

    The first Swedish reactor R1, constructed at the Royal Inst. of Technology in Stockholm, went critical in July 1954. This report presents historical aspects of the reactor, in particular about the reactor as a research instrument and a centre for physical science. The tensions between its role as a prototype and a step in the development of power reactors and that as a scientific instrument are especially focused.

  18. Prototype fast reactor steam generator unit pressure vessel repairs

    International Nuclear Information System (INIS)

    Daniels, B.D.; Green, D.; Henderson, J.D.C.

    1993-01-01

    The prototype fast reactor at Dounreay has experienced a number of unscheduled shutdowns due to leaking reheater and superheater shell welds. There was a need to determine the cracking mechanism and to design a general repair technique simultaneously. Detailed investigations revealed that the crack locations correlated with the positions of rectification welds made at the time of vessel manufacture. A creep crack growth mechanism was identified; this requires through wall residual stress for through cracks to develop. A repair technique has been devised and successfully applied to the sites of a number of leaks. (author)

  19. Upgraded prototype-reactor internal pump for ABWR

    International Nuclear Information System (INIS)

    Kumagai, Mikio; Amemori, Shiro; Saito, Takehiko

    1988-01-01

    In 1983, Toshiba, using their own technology, manufactured a commercial grade reactor internal pump (RIP). Recently, however, a licensing agreement with KSB of West Germany covering the RIP technology, has combined the know-how of KSB with Toshiba's technology to produce a truly high-quality prototype RIP. The pump produces the required coreflow for ABWR at low speed and with high efficiency, and simply by increasing the pump speed to the prior level, the coreflow can be further increased for such advantages as improved fuel cycle economy. Here, the advanced features and test results of the RIP are summarized. (author)

  20. Prototype fuel fabrication for nuclear reactors of Laguna Verde

    International Nuclear Information System (INIS)

    Nocetti, C.; Torres, J.; Medrano, A.

    1996-01-01

    Four prototype fuel bundles for the Laguna Verde Nuclear Power Plant have been fabricated. the type of nuclear fuel produced is described and the process used is commented. As an example of the fabrication criteria adopted, the production model to determine the density of the U O 2 pellets for the different batches of ceramic powder is described. the results are evaluated using the statistical indexes C p and C pk . (author)

  1. Prototype plant for nuclear process heat (PNP)

    International Nuclear Information System (INIS)

    Duerrfeld, R.; Kraut-Giesen, G.

    1982-01-01

    1. Goals: Verification of owner's interests during experimental and engineering phase of nuclear coal gasification. 2. Method: 2.1 Witnessing and evaluating of experimental results from running test facilities. 2.2 Influencing experimental program. 2.3 Participation in important meetings of PNP-project. 3. Results: From present point of view the realization of nuclear coal gasification with a nuclear high temperature reactor (HTR) in accordance with the present technical status as well as meeting the existing safety regulations seems to be feasable. R+D-work will be needed for affirmation of design. The gasification of hard coal basing on the allothermal principal has proved to be possible. The examination of the gasifier on a pilot scale is not yet done. The design work for the pilot plant should be started immediately, particularly keeping in mind the decision for erection of PNP in 1990. The calculation of production costs in comparison to autothermal gasification processes is promising better economics, if uncertainties of investment calculation are deemed to be neglectable. (orig.) [de

  2. Risk-oriented analysis on the German prototype fast breeder reactor SNR-300

    International Nuclear Information System (INIS)

    Bayer, A.; Koeberlein, K.; Gesellschaft fuer Reaktorsicherheit, Garching, Germany)

    1984-01-01

    On request of a fact-finding committee of the German Federal Parliament, a risk-oriented analysis on the SNR-300, the German prototype fast breeder reactor, has been performed to allow a pragmatic safety comparison of the SNR-300 and a modern light-water reactor. Results of the technical plant analysis have been summarized in seven release categories. Accident consequences have been calculated for the actual site at Kalkar/Rhine. The results indicate that for the SNR-300 both the frequency of major accidents and the consequences of accidents are smaller than for the pressurized-water reactor analyzed in the German Risk Study. This article summarizes the methods and main results of the analysis of the SNR-300

  3. Experience of secondary cooling system modification at prototype fast breeder reactor MONJU (Translated document)

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Sakamoto, Yoshihiko

    2010-09-01

    The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident that occurred in December 1995. After the accident, an investigation into the cause and a comprehensive safety review of the plant were conducted, and various countermeasures for sodium leak were examined. Modification work commenced in September 2005. Since sodium, a chemically active material, is used as coolant in MONJU, the modification work required work methods suitable for the handling of sodium. From this perspective, the use of a plastic bag when opening the sodium boundary, oxygen concentration control in a plastic bag, slightly-positive pressure control of cover gas in the systems, pressing and cutting with a roller cutter to prevent the incorporation of metal fillings, etc. were adopted, with careful consideration given to experience and findings from previous modification work at the experimental fast reactor JOYO and plants abroad. Owing to these work methods, the modification work proceeded close to schedule without incident. (author)

  4. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  5. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  6. Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

    International Nuclear Information System (INIS)

    Nishio, S.; Ueda, S.; Kurihara, R.; Kuroda, T.; Miura, H.; Sako, K.; Takase, H.; Seki, Y.; Adachi, J.; Yamazaki, S.; Hashimoto, T.; Mori, S.; Shinya, K.; Murakami, Y.; Senda, I.; Okano, K.; Asaoka, Y.; Yoshida, T.

    2000-01-01

    If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal

  7. Demonstrated operational and inherent safety of the prototype fast reactor (PFR)

    International Nuclear Information System (INIS)

    Smedley, J.A.; Gregory, C.V.; Judd, A.M.

    1983-01-01

    The Prototype Fast Reactor (PFR) is sited at Dounreay, on the north coast of Scotland in the United Kingdom, and has been in operation since 1974. Three aspects of the safety of the reactor are described, including the all-important practical consideration of operational safety, a demonstration of the limited consequences of a sodium/water reaction in a steam generator and the ability of the reactor to protect itself against highly improbable incidents. Attention is drawn to the low radiation levels in the plant and the correspondingly low dose rate to personnel. A feature of PFR operation has been the stable and predictable behaviour of its core together with the high degree of reliability exhibited by the engineered safety system. No failures have occurred within the standard driver charge but two experimental fuel pins suffered cladding failure, which was detected easily by the fission gas and delayed neutron detection systems. In the steam generating units sodium and water are separated by the single steel wall of the steam tubes. Although no under-sodium leak has occurred, an experimental programme is continuing and demonstrates that were any such leak to occur its consequences would be containable and would not result in the release of sodium to the environment or any breach of the reactor containment. The final section describes the inherent safety features of the reactor which enable it to survive a range of very improbable incidents even when the engineered safeguards fail. The features considered are natural circulation, which has been demonstrated by reactor experiment; the reactor's negative power coefficient, which, for example, enables the reactor to survive a complete loss of heat sink; and the durability of the fuel pins, demonstrated by a series of boiling experiments in the Dounreay Fast Reactor (DFR). (author)

  8. Lunar Plants Prototype for Moon Express

    Data.gov (United States)

    National Aeronautics and Space Administration — The goal of our project is to bring the first full life cycle to the moon: to demonstrate germination of plants in lunar gravity and radiation.The Moon Express...

  9. Design of first reactor protection system prototype for C A R E M reactor

    International Nuclear Information System (INIS)

    Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

    2006-01-01

    In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

  10. Role of land-based prototype plants in propulsion nuclear power plants engineering

    International Nuclear Information System (INIS)

    Voronin, V.E.; Prokhorov, Yu.A.

    1993-01-01

    Prototype plants provide a powerful tool for accomplishing tasks of development and construction of newly designed new power plants (NPPs). Leaving aside momentary political or economical considerations, one should admit that the use of prototype plants in testing of new NPPs is quite a necessity. To make the most of prototype plant, its commissioning should precede lead plant construction by 2-3 years. To make good use of prototype plants, a set of basic requirements should be fulfilled: greatest possible identity beteen the facility under test and a new series NPP; provision of high performance data acquisitoin, processing and storage firmware and a modelling system using update computer technique; and developed science infrastructure, engineering support and adequate maintenance. Prototype plants should comply with safety requirements to meet environmental protection standards

  11. Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

    2003-01-01

    Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

  12. Annual technical report of the prototype fast breeder reactor Monju. 2012

    International Nuclear Information System (INIS)

    2013-09-01

    The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2012. From the aspect of the design evaluation, the following items are summarized: 1) Comprehensive safety assessments of Monju taking into account the accident at Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company, 2) Evaluation of nuclear characteristics based on the data of core confirmation test, 3) Evaluation of hydrogen flux from steam generator tubes, 4) Construction of the advanced safeguards system, 5) Development of a plant dynamics analytical model for the Monju ex-vessel fuel storage system. Then, from the aspect of the maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management, 2) Recovery of in vessel transfer machine dropping accident, 3) Work management by introduction of packaged isolation, 4) Evaluation of result of vibration control of RID sampling blower for secondary sodium loop. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

  13. Application of PSA level 1 for the Fugen prototype ATR plant

    International Nuclear Information System (INIS)

    Ikeda, H.; Iguchi, Y.; Sotsu, M.; Seki, O.; Satou, S.

    1997-01-01

    This paper presents application of PSA Level-1 for Prototype Advanced Thermal Reactor Plant ''Fugen'' in consideration of the design characteristics of such reactor and verify the safety aspect of Fugen using thus established procedure, ATR resembles the boiling water reactor (BWR) in a number of points, but there are also some differences between the ATR and the BWR. Therefore. PSA procedure have been established by taking such difference into consideration and by referring to experience of PSA in USA and Japan. Moreover, the core damage frequency was calculated on Fugen by using thus established procedure. As a result, it was verified that results including the maximum value of the uncertainty estimation were found to be quite satisfactory against the target value of reactor damage frequency defined by the International Atomic Energy Agency (IAEA). (author)

  14. Start-up test of the prototype heavy water reactor 'FUGEN', (1)

    International Nuclear Information System (INIS)

    Ando, Hideki; Kawahara, Toshio

    1982-01-01

    The advanced thermal prototype reactor ''Fugen'' is a heavy water-moderated, boiling light water-cooled power reactor with electric output of 165 MW, which has been developed since 1966 as a national project. The start-up test was begun in March, 1978, being scheduled for about one year, and in March, 1979, it passed the final pre-use inspection and began the full scale operation. In this paper, the result of the start-up test of Fugen is reported. From the experience of the start-up test of Fugen, the following matters are important for the execution of start-up test. 1) Exact testing plan and work schedule, 2) the organization to perform the test, 3) the rapid evaluation of test results and the reflection to next testing plan, and 4) the reflection of test results to rated operation, regular inspection and so on. In the testing plan, the core characteristics peculiar to Fugen, and the features of heavy water-helium system, control system and other equipment were added to the contents of the start-up test of BWRs. The items of the start-up test were reactor physics test, plant equipment performance test, plant dynamic characteristic test, chemical and radiation measurement, and combined test. The organization to perform the start-up test, and the progress and the results of the test are reported. (Kako, I.)

  15. The first Swedish nuclear reactor - from technical prototype to scientific instrument

    International Nuclear Information System (INIS)

    Fjaestad, M.

    2001-01-01

    The first Swedish reactor R1, constructed at the Royal Inst. of Technology in Stockholm, went critical in July 1954. This report presents historical aspects of the reactor, in particular about the reactor as a research instrument and a centre for physical science. The tensions between its role as a prototype and a step in the development of power reactors and that as a scientific instrument are especially focused

  16. Design, manufacture and installation of measuring and control equipments for the advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Hirota, Shigeo; Kawabata, Yoshinori

    1979-01-01

    The advanced thermal prototype reactor ''Fugen'' attained the criticality on March 20, 1978, and 100% output operation on November 13, 1978. On March 20, 1979, it passed the final inspection, and since then, it has continued the smooth operation up to now. The measuring and control equipments are provided for grasping the operational conditions of the plant and operating it safely and efficiently. At the time of designing, manufacturing and installing the measuring and control equipments for Fugen, it was required to clarify the requirements of the plant design, to secure the sufficient functions, and to improve the operational process, maintainability and the reliability and accuracy of the equipments. Many design guidelines and criteria were decided in order to coordinate the conditions among five manufacturers and give the unified state as one plant. The outline of the instrumentations for neutrons, radiation monitoring and process data, the control systems for reactivity, reactor output, pressure and water supply, the safety protection system, and the process computer are described. Finally, the installations and tests of the measuring and control equipments are explained. The aseismatic capability of the equipments was confirmed. (Kako, I.)

  17. Annual technical report of the prototype fast breeder reactor Monju. 2013

    International Nuclear Information System (INIS)

    2014-08-01

    The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the principal achievements and the data related to the plant management in Monju in fiscal 2013. From the aspect of the management and design evaluation, the following items are summarized: 1) Current status of coping with the order from NRA to alter the safety regulations. 2) Implementation status of reformation of Monju. 3) Current status of the additional geological surveys of crush zones at the Monju site. 4) Development of core seismic assessment method for FBR. Then, from the aspect of the operation and maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management (Part 2). 2) Performance confirmation of the failed fuel detection and location system. 3) Deviation from the limiting conditions for operation in the emergency diesel generator periodic test and so on. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

  18. Multiple recycling of fuel in prototype fast breeder reactor in a closed ...

    Indian Academy of Sciences (India)

    Our previous study in this regard for the prototype fast breeder reactor ... This study aims at finding the feasibility of multiple recycling of PFBR fuel with external ...... maximum allowable Pu content in fuel based on chemistry/metallurgical ...

  19. Regulatory requirements for desalination plant coupled with nuclear reactor plant

    International Nuclear Information System (INIS)

    Yune, Young Gill; Kim, Woong Sik; Jo, Jong Chull; Kim, Hho Jung; Song, Jae Myung

    2005-01-01

    A small-to-medium sized reactor has been developed for multi-purposes such as seawater desalination, ship propulsion, and district heating since early 1990s in Korea. Now, the construction of its scaled-down research reactor, equipped with a seawater desalination plant, is planned to demonstrate the safety and performance of the design of the multi-purpose reactor. And the licensing application of the research reactor is expected in the near future. Therefore, a development of regulatory requirements/guides for a desalination plant coupled with a nuclear reactor plant is necessary for the preparation of the forthcoming licensing review of the research reactor. In this paper, the following contents are presented: the design of the desalination plant, domestic and foreign regulatory requirements relevant to desalination plants, and a draft of regulatory requirements/guides for a desalination plant coupled with a nuclear reactor plant

  20. High temperature alloys for the primary circuit of a prototype nuclear process heat plant

    International Nuclear Information System (INIS)

    Ennis, P.J.; Schuster, H.

    1979-01-01

    As part of a comprehensive materials test programme for the High Temperature Reactor Project 'Prototype Plant for Nuclear Process Heat' (PNP), high temperature alloys are being investigated for primary circuit components operating at temperatures above 750 0 C. On the basis of important material parameters, in particular corrosion behaviour and mechanical properties in primary coolant helium, the potential of candidate alloys is discussed. By comparing specific PNP materials data with the requirements of PNP and those of conventional plant, the implications for the materials programme and component design are given. (orig.)

  1. Operating performance of the prototype heavy water reactor Fugen

    International Nuclear Information System (INIS)

    1984-01-01

    Since the full scale operation was started in March, 1979, the ATR Fugen power station has been verifying the performance and reliability of the machinery and equipment, uranium-plutonium mixed oxide fuel and so on, and obtaining the technical prospect for putting ATRs in practical use by accumulating operation and maintenance techniques, through about five years of operation. In this report, the operational results of the Fugen power station are described. Fugen is a heavy water-moderated, boiling light water-cooled, pressure tube type reactor with 165 MWe output. As of the end of March, 1984, the total generated electric power was about 4.3 billion kWh, and the operation time was about 27,000 hours. The mean capacity ratio reached 58.8%. During the operation period, troubles including plant shutdown occurred eight times, but generally the performance and reliability of the machinery and equipment have been good. 580 fuels including 284 MOX fuels have been charged, but fuel breaking did not occur at all. The consumption of heavy water and the leak of tritium did not cause problem. The management of the core and fuel, the management of maintenance, the quality control of cooling water and heavy water, radiation control and the management of wastes are reported. (Kako, I.)

  2. Reactor plant for Belene NPP completion

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

    2004-01-01

    Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

  3. R&D on high-power dc reactor prototype for ITER poloidal field converter

    Energy Technology Data Exchange (ETDEWEB)

    Li, Chuan [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Song, Zhiquan; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Zhang, Ming, E-mail: zhangming@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yu, Kexun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Qin, Xiuqi [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China)

    2015-10-15

    Highlights: • A new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented. • Theoretical analysis, finite-element simulation and prototype test verification are applied on the design. • The results of temperature rise and transient fault current test of prototypes are introduced and analyzed. • The success of tests demonstrates that the proposed structure is of high reliability and availability. - Abstract: This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

  4. Rapid Control Prototyping Plataform for Didactic Plant Motor DC

    Directory of Open Access Journals (Sweden)

    Cristian Bazán-Orobio

    2013-06-01

    Full Text Available In this paper a design, implementation and validation of a Rapid Control Prototype platform for a plant based on a DC motor is proposed. This low-cost prototype provides of an electronic card (with a motor DC and sensors manipulated by PC with free software tools using Linux, Scilab / Scicos and RTAI-Lab. This RCP System allows developing speed -position control trainings by using different types of PID industrial controllers with anti – wind up and bump less transfer schemes. We develop a speed control application structured in four steps: identification, controller design, simulation and real time control, where there are pedagogical advantages of a platform that not only allows simulation but also real-time control of a plant.

  5. Engineering Design of ITER Prototype Fast Plant System Controller

    Science.gov (United States)

    Goncalves, B.; Sousa, J.; Carvalho, B.; Rodrigues, A. P.; Correia, M.; Batista, A.; Vega, J.; Ruiz, M.; Lopez, J. M.; Rojo, R. Castro; Wallander, A.; Utzel, N.; Neto, A.; Alves, D.; Valcarcel, D.

    2011-08-01

    The ITER control, data access and communication (CODAC) design team identified the need for two types of plant systems. A slow control plant system is based on industrial automation technology with maximum sampling rates below 100 Hz, and a fast control plant system is based on embedded technology with higher sampling rates and more stringent real-time requirements than that required for slow controllers. The latter is applicable to diagnostics and plant systems in closed-control loops whose cycle times are below 1 ms. Fast controllers will be dedicated industrial controllers with the ability to supervise other fast and/or slow controllers, interface to actuators and sensors and, if necessary, high performance networks. Two prototypes of a fast plant system controller specialized for data acquisition and constrained by ITER technological choices are being built using two different form factors. This prototyping activity contributes to the Plant Control Design Handbook effort of standardization, specifically regarding fast controller characteristics. Envisaging a general purpose fast controller design, diagnostic use cases with specific requirements were analyzed and will be presented along with the interface with CODAC and sensors. The requirements and constraints that real-time plasma control imposes on the design were also taken into consideration. Functional specifications and technology neutral architecture, together with its implications on the engineering design, were considered. The detailed engineering design compliant with ITER standards was performed and will be discussed in detail. Emphasis will be given to the integration of the controller in the standard CODAC environment. Requirements for the EPICS IOC providing the interface to the outside world, the prototype decisions on form factor, real-time operating system, and high-performance networks will also be discussed, as well as the requirements for data streaming to CODAC for visualization and

  6. Nuclear power plant with several reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

    1972-05-10

    A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

  7. ITER Fast Plant System Controller prototype based on PXIe platform

    International Nuclear Information System (INIS)

    Ruiz, M.; Vega, J.; Castro, R.; Sanz, D.; López, J.M.; Arcas, G. de; Barrera, E.; Nieto, J.; Gonçalves, B.; Sousa, J.; Carvalho, B.; Utzel, N.; Makijarvi, P.

    2012-01-01

    Highlights: ► Implementation of Fast Plant System Controller (FPSC) for ITER CODAC. ► Efficient data acquisition and data movement using EPICS. ► Performance of PCIe technologies in the implementation of FPSC. - Abstract: The ITER Fast Plant System Controller (FPSC) is based on embedded technologies. The FPSC will be devoted to both data acquisition tasks (sampling rates higher than 1 kHz) and control purposes (feedback loop actuators). Some of the essential requirements of these systems are: (a) data acquisition and data preprocessing; (b) interfacing with different networks and high speed links (Plant Operation Network, timing network based on IEEE1588, synchronous data transference and streaming/archiving networks); and (c) system setup and operation using EPICS (Experimental Physics and Industrial Control System) process variables. CIEMAT and UPM have implemented a prototype of FPSC using a PXIe (PCI eXtension for Instrumentation) form factor in a R and D project developed in two phases. The paper presents the main features of the two prototypes developed that have been named alpha and beta. The former was implemented using LabVIEW development tools as it was focused on modeling the FPSC software modules, using the graphical features of LabVIEW applications, and measuring the basic performance in the system. The alpha version prototype implements data acquisition with time-stamping, EPICS monitoring using waveform process variables (PVs), and archiving. The beta version prototype is a complete IOC implemented using EPICS with different software functional blocks. These functional blocks are integrated and managed using an ASYN driver solution and provide the basic functionalities required by ITER FPSC such as data acquisition, data archiving, data pre-processing (using both CPU and GPU) and streaming.

  8. Study concerning the erection within the precincts of INR Pitesti of TRIGA prototype nuclear heating plant

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Ionescu, M.; Constantin, L.

    1993-01-01

    This paper presents the problems of nuclear plant energy production as heating source for industrial processes and urban district heating. The study is based on the TRIGA concept due to some of its advantages in comparison with other concepts. The system solutions for a prototype implementation and the aspects of the economical and financial efficiency are outlined. The conclusion is drawn that the TRIGA 53 MWt-reactor is suitable to meet the heating needs of urban and industrial heating systems in this country

  9. Prototype thermochemical heat storage with open reactor system

    NARCIS (Netherlands)

    Zondag, H.A.; Kikkert, B.; Smeding, S.F.; Boer, de R.; Bakker, M.

    2013-01-01

    Thermochemical (TC) heat storage is an interesting technology for future seasonal storage of solar heat in the built environment. This technology enables high thermal energy storage densities and low energy storage losses. A small-scale laboratory prototype TC storage system has been realized at

  10. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder ...

  11. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    International Nuclear Information System (INIS)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option

  12. Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

    International Nuclear Information System (INIS)

    Ishiyama, Shintaro; Muto, Yasushi; Kato, Yasuyoshi; Nishio, Satoshi; Hayashi, Takumi; Nomoto, Yasunobu

    2008-01-01

    Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO 2 (S-CO 2 ) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480degC, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO 2 turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO 2 cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m 3 and 7240 m 3 for the steam turbine system and S-CO 2 turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO 2 than in H 2 O. Therefore, the S-CO 2 turbine system is recommended to the fusion reactor system than the steam turbine system. (author)

  13. Annual technical report of the prototype fast breeder reactor Monju. 2011

    International Nuclear Information System (INIS)

    2012-08-01

    The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2011. From the aspect of the design evaluation, the following items are summarized: 1) the evaluation of the decay heat removal of Monju core by natural convection, and the safety measures against earthquake and tsunami, which were carried out from the lessons learned at the Fukushima-daiichi accident due to the Great East Japan Earthquake on March 11, 2011, 2) the control rod worth confirmation and the evaluation of nuclear data library based on the data of Core Confirmation Test, which is the first step of Monju system startup test restarted in 2010, 3) the evaluation of the hydrogen concentration behavior, which detects the leak of water from the heat transfer tube of steam generator. Then, from the aspect of the maintenance technology, the following items are summarized: 1) the results of the function confirmation test on the water/steam system, after the long-term suspension, 2) confirmation of the integrity of cracked cylinder liners of emergency diesel generator, 3) replacement of the annulus ventilation duct, 4) evaluation of reduction of the periodic inspection schedule after full power operation. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

  14. Plant maintenance and advanced reactors, 2006

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal (ed.)

    2006-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

  15. Prototype equipment status monitor for plant operational configuration management

    International Nuclear Information System (INIS)

    DeVerno, M.; Trask, D.; Groom, S.

    1998-01-01

    CANDU plants, such as the Point Lepreau GS, have tens of thousands of operable devices. The status of each operable device must be immediately available to plan and execute future changes to the plant. Historically, changes to the plant's operational configuration have been controlled using manual and administrative methods where the status of each operable device is maintained on operational flowsheets located in the work control area of the main control room. The operational flowsheets are used to plan and develop Operating Orders (OOs) or Order-to-Operate (OTOs) and the control centre work processes are used to manage their execution. After performing each OO procedure, the operational flowsheets are updated to reflect the new plant configuration. This process can be very time consuming, and due to the manual processes, can lead to the potential for time lags and errors in the recording of the current plant configuration. Through a cooperative research and development program, Canadian CANDU utilities and Atomic Energy of Canada Limited, the design organization, have applied modern information technologies to develop a prototype Equipment Status Monitor (ESM) to address processes and information flow for efficient operational configuration management. The ESM integrates electronic operational flowsheets, equipment databases, engineering and work management systems, and computerized procedures to assess, plan, execute, track, and record changes to the plant's operational configuration. This directly leads to improved change control, more timely and accurate plant status information, fewer errors, and better decision making regarding future changes. These improvements to managing the plant's operational configuration are essential to increasing plant safety, achieving a high plant availability, and maintaining high capability and capacity factors. (author)

  16. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    Mokhov, V.A.

    2010-01-01

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  17. Core disruptive accident analysis in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

    2002-01-01

    Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

  18. Sodium tests on an integrated purification prototype for a sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Abramson, R.

    1984-04-01

    This paper describes sodium tests performed on the integrated primary sodium purification prototype of the Creys Malville Super Phenix 1 fast breeder reactor. These tests comprised: - hydrostatic test, - thermal tests, - handling tests. They enabled a number of new technological arrangements to be qualified and provided the necessary information for the design and construction of the Super Phenix 1 purification units

  19. Reactivity monitoring for safety purposes on the UK prototype fast reactor

    International Nuclear Information System (INIS)

    Lord, D.J.; Wilkes, D.J.

    1987-01-01

    The small size and high rating of the liquid metal cooled fast breeder reactor (LMFBR) make the provision of safety related instrumentation for individual subassemblies both difficult and expensive. Global monitoring of the core is thus very attractive. Reactivity monitoring is an important part of such global monitoring. Reactivity monitoring on a short timescale (a few seconds) is used on the UK Prototype Fast Reactor (PFR) as a trip parameter and long-term reactivity monitoring is being developed as a means of providing early warning of slowly developing faults. Results are presented from PFR to demonstrate the capabilities of reactivity monitoring in an operational fast reactor power station. (author)

  20. Total decay heat estimates in a proto-type fast reactor

    International Nuclear Information System (INIS)

    Sridharan, M.S.

    2003-01-01

    Full text: In this paper, total decay heat values generated in a proto-type fast reactor are estimated. These values are compared with those of certain fast reactors. Simple analytical fits are also obtained for these values which can serve as a handy and convenient tool in engineering design studies. These decay heat values taken as their ratio to the nominal operating power are, in general, applicable to any typical plutonium based fast reactor and are useful inputs to the design of decay-heat removal systems

  1. Construction work for prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Suzuki, Yasuyuki; Tsuji, Koichi; Shimizu, Hisashi

    1991-01-01

    The main construction work of MONJU was started from the excavation for building foundation in October 1985, the containment vessel was prepared in April 1987, the reactor vessel was installed in October 1988, and the installation of the whole equipment was completed in April 1991. Fuji Electric made preparations for construction matching the above master schedule in consideration of construction schedule quality assurance and safety and accomplished the work within the scheduled time without personal injury as long as 2.35 million hours. (author)

  2. Advanced liquid metal reactor plant control system

    International Nuclear Information System (INIS)

    Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

    1993-01-01

    The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

  3. The remaining risk to be accepted with test facilities and prototype plants, and the relevant legal provisions of nuclear law

    International Nuclear Information System (INIS)

    Mayinger, T.

    1995-01-01

    The first chapter explains the provisions laid down in nuclear law to assure that precaution is taken to prevent damage resulting from the operation of nuclear power reactors, in order to set a line for comparison with the relevant legal provisions relating to test facilities and prototype plants. The comparative analysis shows that the means and methods of precaution are defined to comprise three approaches, namely measures taken to avert danger, measures taken to prevent danger, and measures for (remaining) risk minimization. All three approaches are intended to prevent occurrence of specifically nuclear events. The second chapter characterizes power reactors, prototype plant and test facilities and develops criteria for distinction. The third chapter establishes the systematics for comparison, showing whether and how the mandatory precaution to prevent damage defined for power reactors, prototype plant, and test facilities can be distinguished from each other, the results being represented in a systematic survey of licensing requirements as laid down in section 7, sub-section 2 ATG (Atomic Energy Act). (orig./HP) [de

  4. Diagnostics in the hostile environments of a prototype fusion reactor

    International Nuclear Information System (INIS)

    Osher, J.E.

    1982-01-01

    Various facets of a thermonuclear type plasma that will likely require special considerations or hardening of applied diagnostic instrumentation are reviewed. The discussion will include both on-line diagnostic instrumentation requirements for satisfactory operation and considerations to reduce integrated radiation damage sufficiently for a reasonable diagnostic lifetime. Several new diagnostics aimed specifically at measurements of the plasma characteristics most appropriate to a thermonculear reactor type plasma are discussed. This will include instrumentation needed to make quantitative energy flow measurements during long term operation with the expected high input power sources, and locally very high wall power loadings. The second part of this lecture will broaden diagnostics to include materials damage measurements needed for engineering design studies. This will include needed diagnostic instrumentation to assess first wall damage, sputtering erosion at walls (and high power beam dumps), and radiation damage to components such as insulators

  5. Study of the accidental risk of the German fast breeder prototype reactor SNR-300

    International Nuclear Information System (INIS)

    Koeberlein, K.

    1983-01-01

    A fact-finding committee of the German Federal Parliament in July 1980 recommended to perform a 'risk-oriented study' of the SNR-300, the German 300 MWe fast breeder prototype reactor being under construction at Kalkar. The main aim of this study was to allow a comparative safety evaluation between the SNR-300 and a modern PWR, thus to prepare a basis for a political decision on the SNR-300. Methods and main results of the study are presented in this paper. In the first step of the study six groups of accidents have been identified which may initiate core destruction. By reliability analyses, expected frequency of each group has been estimated. Conditional probabilities for conceivable reactor tank failure modes have been analysed. Tank failure after core destruction leads to release of energy and radioactive material into the containment system. Such accident sequences have been pursued further. Based on a number of core destruction initiators and tank failure modes and various combinations of success and failure states of the containment systems, detailed calculations of different containment scenarious were carried out. From the results of the plant systems analysis, five release categories have been defined. Possible effects of external events and releases of radioactivity from the spent fuel storage pool have also been analysed. In order to quantify the degree of uncertainty of the calculated frequencies, subjective probability distributions of fixed, but inaccurately known quantities have been propagated through the calculations. Using the release categories as input, accident consequences were calculated for the site Kalkar. Though the uncertainty bandwidths for the accident frequencies estimated in the SNR-300 analysis are much wider than for the PWR, the analysis indicates that frequencies of severe accidents, and consequences, are smaller for the SNR-300 than for the PWR as analysed in the 'German Risk Study'. (orig.)

  6. A Takagi-Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Yue [Institute of Nuclear and New Energy Technology, Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing (China); Coble, Jamie [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S) fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

  7. A Takagi–Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

    Directory of Open Access Journals (Sweden)

    Yue Yuan

    2017-08-01

    Full Text Available Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

  8. Discharges from a fast reactor reprocessing plant

    International Nuclear Information System (INIS)

    Barnes, D.S.

    1987-01-01

    The purpose of this paper is to assess the environmental impact of the calculated routine discharges from a fast reactor fuel reprocessing plant. These assessments have been carried out during the early stages of an evolving in-depth study which culminated in the design for a European demonstration reprocessing plant (EDRP). This plant would be capable of reprocessing irradiated fuel from a series of European fast reactors. Cost-benefit analysis has then been used to assess whether further reductions in the currently predicted routine discharges would be economically justified

  9. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  10. Overall plant concept for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Yamaki, Hideo; Davies, S.M.; Goodman, L.

    1984-01-01

    Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

  11. Post-accident monitoring systems in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Suriya Murthy, N.; Sivasailanathan, Vidhya; Ananth, Allu; Roy, Kallol

    2018-01-01

    PFBR is a 500 MW(e) MOX fueled and sodium cooled fast reactor (SFR) under advanced stage of commissioning at Kalpakkam. Currently, the main vessel is preheated and sodium has been charged into two secondary loops that are operated in recirculation mode. In order to characterize the radiation field and contamination, the workplace monitoring is undertaken using installed monitors that are commissioned and made operational. This helps to ensure radiological protection during normal operating conditions. On the other hand, radiological monitoring in emergency conditions is quite different. For undertaking the mitigative accident management, a set of specialized nuclear instruments called post-accident monitoring systems (PAMS) which include radiation monitors are stipulated. The Fukushima Daiichi accident emphasized the importance and need for reliable accident monitoring instrumentation to indicate the safety functions during the progression and aftermath of accident in NPP. In PFBR, the PAMS are integrated with other monitoring systems in design stage itself to manage the measurements and indicating the safety functions for implementing EOP and SAMG

  12. Lesson Learned in Preparation for Decommissioning of Three Canadian Prototype Power Reactors

    International Nuclear Information System (INIS)

    Vickerd, Meggan; Kenny, Stephen

    2016-01-01

    Lesson learned by Canadian Nuclear Laboratories (CNL)(former AECL) in preparation for decommissioning of three Prototype Reactors is a result of various strategies used for each site. CNL is responsible for the eventual decommissioning of three prototype power reactors; Nuclear Power Demonstration (NPD), Gentilly-1 and Douglas Point. Each of the Canadian prototype power reactor sites shutdown using different strategies. Depending on the site location, configuration, and intended designation of the respective sites, the individual facility systems (ventilation, electrical system, fire detection etc.) were also shut down using different strategies and operating objectives. As CNL embarks on decommissioning the first Canadian prototype reactor, this paper will reflect on the lessons learned over the past thirty years and what CNL is adjusting in the decommissioning strategy to prepare better plans for the future. The Nuclear Power Demonstration Nuclear Generating Station (NPDNGS) was constructed in late 1950's and operated from 1962 to 1987 when it was permanently shutdown after exceeding its operational goals. The NPD reactor was the first Canadian nuclear power reactor and it consisted of a single 20 MWe pressurized heavy water reactor located on a single facility site in Rolphton, Ontario. The NPD facility was shutdown to a 'Cold, Dark and Quiet' state and is maintained using an unmanned strategy by managing the site remotely with active fire detection and security surveillance systems, minimal electrical supply and an active ventilation system which is operated periodically to allow for intermittent inspections. The Douglas Point Nuclear Generating Station (DPNGS) was constructed in the early 1960's and operated from 1968 to 1984 when it was permanently shutdown. It consisted of a 200 MW prototype Canada Deuterium Uranium (CANDU) reactor and is embedded on the Bruce Power site near Kincardine, Ontario. The Douglas Point site is maintained in a

  13. Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, A; Bowden, N; Misner, A; Palmer, T

    2007-06-27

    In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to 3.5% within 7 days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

  14. CO2 Reduction Assembly Prototype Using Microlith-Based Sabatier Reactor for Ground Demonstration

    Science.gov (United States)

    Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

    2014-01-01

    The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA's cabin Atmosphere Revitalization System (ARS) and In-Situ Resource Utilization (ISRU) architectures for both low-earth orbit and long-term manned space missions. Carbon dioxide can be reacted with H2, obtained from the electrolysis of water, via Sabatier reaction to produce methane and H2O. Methane can be stored and utilized as propellant while H2O can be either stored or electrolyzed to produce oxygen and regain the hydrogen atoms. Depending on the application, O2 can be used to replenish the atmosphere in human-crewed missions or as an oxidant for robotic and return missions. Precision Combustion, Inc. (PCI), with support from NASA, has previously developed an efficient and compact Sabatier reactor based on its Microlith® catalytic technology and demonstrated the capability to achieve high CO2 conversion and CH4 selectivity (i.e., =90% of the thermodynamic equilibrium values) at high space velocities and low operating temperatures. This was made possible through the use of high-heat-transfer and high-surface-area Microlith catalytic substrates. Using this Sabatier reactor, PCI designed, developed, and demonstrated a stand-alone CO2 Reduction Assembly (CRA) test system for ground demonstration and performance validation. The Sabatier reactor was integrated with the necessary balance-of-plant components and controls system, allowing an automated, single "push-button" start-up and shutdown. Additionally, the versatility of the test system prototype was demonstrated by operating it under H2-rich (H2/CO2 of >4), stoichiometric (ratio of 4), and CO2-rich conditions (ratio of <4) without affecting its performance and meeting the equilibrium-predicted water recovery rates. In this paper, the development of the CRA test system for ground demonstration will be discussed. Additionally, the performance results from testing the system at

  15. Improving plant availability by predicting reactor trips

    International Nuclear Information System (INIS)

    Frank, M.V.; Epstein, S.A.

    1986-01-01

    Management Ahnalysis Company (MAC) has developed and applied two complementary software packages called RiTSE and RAMSES. Together they provide an mini-computer workstation for maintenance and operations personnel to dramatically reduce inadvertent reactor trips. They are intended to be used by those responsible at the plant for authorizing work during operation (such as a clearance coordinator or shift foreman in U.S. plants). They discover and represent all components, processes, and their interactions that could case a trip. They predict if future activities at the plant would cause a reactor trip, provide a reactor trip warning system and aid in post-trip cause analysis. RAMSES is a general reliability engineering software package that uses concepts of artificial intelligence to provide unique capabilities on personal and mini-computers

  16. Design safety limits in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Puthiyavinayagam, P.; Roychowdhury, D.G.; Govindarajan, S.; Chellapandi, P.; Singh, Om Pal; Chetal, S.C.

    2002-01-01

    Full text: PFBR is designed to operate at 450 W/cm peak linear heat rating to a peak burn up of 100,000 MWd/t which corresponds to a damage dose of 85 dpa. The targetted reliability is to restrict pin failure to 1 in 10,000. All the design basis events are classified into four categories. Design safety limits imposed for DBE are in terms of temperatures, radiation doses and structural design parameters. Radiation limits are imposed in relation to RCB from the plant personnel and public point of view. Fuel pin integrity is assured with a detailed damage analysis by adopting cumulative damage concept for fixing clad temperature limits. Fuel temperatures are limited to melting point to preclude fuel slumping for events up to category 3. Partial melting is allowed for events in category 4 and the results obtained from transients experiments show that partial melting up to 50% of pellet area does not result in clad failure. Coolant temperatures are limited to boiling point to avoid burnout and reactivity effects

  17. Prototypic Enhanced Risk Monitor Framework and Evaluation - Advanced Reactor Technology Milestone: M3AT-15PN2301054

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Veeramany, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bonebrake, Christopher A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ivans, William J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coles, Garill A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coble, Jamie B. [Univ. of Tennessee, Knoxville, TN (United States); Liu, X. [Univ. of Tennessee, Knoxville, TN (United States); Wootan, David W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mitchell, Mark R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Brass, Mary F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-09-24

    This research report summaries the development and evaluation of a prototypic enhanced risk monitor (ERM) methodology (framework) that includes alternative risk metrics and uncertainty analysis. This updated ERM methodology accounts for uncertainty in the equipment condition assessment (ECA), the prognostic result, and the probabilistic risk assessment (PRA) model. It is anticipated that the ability to characterize uncertainty in the estimated risk and update the risk estimates in real time based on equipment condition assessment (ECA) will provide a mechanism for optimizing plant performance while staying within specified safety margins. These results (based on impacting active component O&M using real-time equipment condition information) are a step towards ERMs that, if integrated with AR supervisory plant control systems, can help control O&M costs and improve affordability of advanced reactors.

  18. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

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

  19. Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Kapoor, K.; Prahlad, B.

    2012-01-01

    Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

  20. Passive and engineered safety features of the prototype fast reactor (PFR), Dounreay

    International Nuclear Information System (INIS)

    Gregory, C.V.

    1991-01-01

    Prototype fast reactor (PFR) combines passive and engineered safety features. Natural convection, a strong negative power coefficient, the decay heat removal system, and a fuel design able to operate beyond failure are all inherent and passive safety features of the PFR. The reliable shutdown system and the protection provided against SGU leaks are example of engineered protection. Experience at PFR demonstrates the worth and potential of a range of passive and engineered safeguards

  1. Long Term Storage with Surveillance of Canadian Prototype Nuclear Power Reactors

    International Nuclear Information System (INIS)

    Janzen, Rick

    2008-01-01

    Atomic Energy of Canada (AECL) was originally formed by the government of Canada in 1952 to perform research in radiation and nuclear areas. In the mid 1950's Canada decided to limit itself to peaceful uses of nuclear energy and AECL embarked on several research and development programs, one of them being the development of nuclear power plants. This led to the development of the CANDU TM design of heavy water power reactors, of which there are now 29 operating around the world. This presentation discusses the present state of the first two CANDU TM prototype reactors and a prototype boiling light water reactor and lessons learnt after being in a long-term storage with surveillance state for more than 20 years. AECL facilities undergo decommissioning by either a prompt or a deferred removal approach. Both approaches are initiated after an operating facility has been declared redundant and gone through final operational shutdown. For the deferred approach, initial decommissioning activities are performed to put the facility into a sustainable, safe, shutdown state to minimize the hazards and costs of the ensuing extended storage with surveillance (SWS) or Safestor phase. At the appropriate time, the facility is dismantled and removed, or put into a suitable condition for re-use. AECL has a number of facilities that were built during its history, and some of these are now redundant or will become redundant in the near future. The deferred removal approach is part of AECL's decommissioning strategy for several reasons: 1. Reduction in radiation doses to workers during the final dismantling, 2. No facilities are available yet in Canada for the management of quantity of wastes arising from decommissioning, 3. Financial constraints presented by the number of facilities that will undergo decommissioning, compared to the availability of funds to carry out the work. This has led to the development of a comprehensive decommissioning plan that includes all of AECL's redundant

  2. 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

  3. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design

  4. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Buden, D.; Ranken, W.A.; Koenig, D.R.

    1980-01-01

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

  5. Generation IV reactors and the ASTRID prototype: lessons from the Fukushima accident

    International Nuclear Information System (INIS)

    Gauche, F.

    2012-01-01

    In France, the ASTRID prototype is an industrial demonstrator of a sodium-cooled fast neutron reactor (SFR), fulfilling the criteria for Generation IV reactors. ASTRID will meet safety requirements as stringent as for third generation reactors, and it takes into account lessons from the Fukushima accident. The objectives are to reinforce the robustness of the safety demonstration for all safety functions. ASTRID will feature an innovative core with a negative sodium void coefficient, it will take advantage of the large thermal inertia of SFR for decay heat removal, and will provide for a design either eliminating the sodium-water reaction, or guaranteeing no consequences for safety in case such reaction would take place. (author)

  6. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  7. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  8. Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Kawahara, Toshio; Matsushita, Tadashi

    1977-01-01

    The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

  9. Development of digital plant protection system for Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Suk-Joon Park

    1998-01-01

    A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author)

  10. Development of digital plant protection system for Korean Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suk-Joon [NSSS Engineering and Development Division, Korea Power Engineering Company, Taejon (Korea, Republic of)

    1998-10-01

    A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author) 8 refs, 4 figs, 3 tabs

  11. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  12. Prestressed reactor vessel for nuclear power plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1982-01-01

    With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

  13. Nuclear reactor plants and control systems therefor

    International Nuclear Information System (INIS)

    de Boer, G.A.; de Hex, M.

    1976-01-01

    A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

  14. Advanced light water reactor plant

    International Nuclear Information System (INIS)

    Giedraityte, Zivile

    2008-01-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  15. BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G. [N.A. Dollezhal Institute ' NIKIET' , PO Box 788, Moscow, 101000 (Russian Federation)

    2006-07-01

    This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with {beta}{sub eff}, low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

  16. BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

    International Nuclear Information System (INIS)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G.

    2006-01-01

    This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with β eff , low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

  17. Pressurised-water reactor plant

    International Nuclear Information System (INIS)

    Bergloeff, D.

    1976-01-01

    This additional patent improves the plant described in main patent 2244562 where the primary coolant pump is built into the housing of the steam raising unit, by overcoming certain difficulties in the transition from the specially extensive tube floor of the steam raising unit to the narrow crossection of the pump penetration. This is achieved by the formation of an unsymmetrical inlet funnel between the tube floor of the steam raising unit and the suction side of the pump. A side wall of the funnel forms the separating wall between the inlet and the outlet chamber below the tube floor of the steam raising unit. The crossection of the inlet funnel is reduced towards the pump by 2/3 to 1/3. Even laminar flow conditions are obtained at all flow velocities. Erosion, which disturbs flow and causes cavitation, no longer occurs. (ORU) [de

  18. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  19. Assessment of weld joints of steam generator of prototype fast breeder reactor by microfocal radiography

    International Nuclear Information System (INIS)

    Venkatraman, B.; Saravanan, T.; Jayakumar, T.; Kalyanasundaram, P.; Raj, B.

    2004-01-01

    The tube to tubesheet (TTS) welds of steam generator of Prototype Fast Breeder Reactor (PFBR) are quite critical. Sodium flows on shell side and water on tube side. Any failure would thus be catastrophic. Apart from defects such as porosities, wall thinning due to concavity is endemic in such joints and needs to be detected. This paper presents the methodologies developed for quantitative evaluation of defects including wall thinning due to concavity in the TTS welds by micro focal radiography. The method has been successfully adopted in the shop floor for the evaluation of TTS welds of steam generator and evaporator. (author)

  20. Applications of artificial intelligence to reactor and plant control

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1989-01-01

    Potential improvements in plant efficiency and reliability are often cited as reasons for developing and applying artificial intelligence (AI) techniques, principally expert systems, to the control and operation of nuclear reactors. Nevertheless, there have been few such applications and then mostly at the prototype level. Therefore, if AI techniques are to contribute to process control, methods must be identified by which rule-based and analytic approaches can be merged. This hypothesis is the basic premise of this article. Presented below are 1. a brief review of the human approach towards process control, 2. a discussion of the suitability of AI methodologies for the performance of control tasks, 3. examples of AI applications to both open- and closed-loop control, 4. an enumeration of unresolved issues associated with the use of AI for control, and 5. a discussion of the possible role of expert system techniques in process control. (orig./GL)

  1. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  2. Plant with nuclear reactor, in particular a thermal reactor

    International Nuclear Information System (INIS)

    Straub, H.

    1988-01-01

    The reactor core of the plant has tubular and vertically movable control rods moved by a flow of coolant under pressure. Each control rod surrounds a similarly tubular guide rod, stationary relative to the reactor core, leaving an annular slot-like space therebetween. The inside of each guide rod forms a first pressure chamber supplied with the coolant under pressure. The upper end of each control rod is closed and has a vertical shaft that extends into the inside of the guide rod and forms therewith a second annular slot-like space. At least one first restriction is provided in the first annular slot-like space and at least one second restriction is provided in the second annular slot-like space. A second pressure chamber is formed between both restrictions. The coolant supplied to the guide rod thus returns to the pressure vessel surrounding the reactor core through the second annular slot-like space, the second pressure chamber and the first annular slot-like space. Controlling means are provided, with which pressure thrusts can be generated if necessary in the coolant within the first pressure chamber. (author) 5 refs., 10 figs

  3. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Calabrese, Carlos R.

    2001-01-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  4. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  5. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  6. Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.; Keddar, A.

    1982-08-01

    The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the 240 Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies

  7. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

    2003-01-01

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

  8. Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Hess, J. D.; Gauld, I. C.; Gulliford, J.; Hill, I.; Okajima, S.

    2017-01-01

    Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Project (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.

  9. Fuel handling grapple for nuclear reactor plants

    International Nuclear Information System (INIS)

    Rousar, D.L.

    1992-01-01

    This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

  10. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  11. RESI-1 and RESI-2: pPrototypes of an information system on reactor safety

    International Nuclear Information System (INIS)

    Schultheiss, G.F.; Eglin, W.; Katz, F.W.; Krings, T.; Pee, A.; Schlechtendahl, E.G.

    1975-04-01

    To demonstrate by practical experience the feasibility of the information system elaborated in the 'Study of an Information System on Reactor Safety RESI' (KFK 1900), the prototype systems RESI-1 and RESI-2 were developed and tested in operation. The two systems have been considerably reduced both in extent and contents as compared to the information system described in the study. The RESI-1 prototype system is a paper version established for verification of all the individual functions before passing over to the computer-aided interactive version RESI-2. RESI-2 is based on the GOLEM system of Siemens. Both protoype systems have proved that the essential features: 1) documentation, 2) formulation of and answering to safety questions, which are relevant with respect to particular licensing cases, 3) formulation of safety questions related to individual reactor types can be managed satisfactorily. All the functions of information retrieval have been tested carefully over several months. Particularities of project development and of the methods elaborated are described in detail and presented in this report. (orig.) [de

  12. A capital cost reduction study on the fast breeder reactor plant

    International Nuclear Information System (INIS)

    Taniyama, H.; Kamei, M.; Moriyama, M.

    1991-01-01

    A capital cost reduction study has been performed for large fast breeder reactor designs. The primary objective of this study is to show a trend of capital cost reduction between FBR plants at the prototype stage, the demonstration stage, and the future commercialization stage. For the FBR plant at the demonstration stage a construction cost comparison with a light water reactor has also been performed, and the target cost of FBR of below 1.5 times that of the light water reactor cost was achieved. To extend the capital cost reduction study, a feasibility study was made to achieve a capital cost of an FBR less than that of a light water reactor. The recommended design is shown as a future commercialization FBR design concept. (author)

  13. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  14. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  15. The Concept of the Use of the Marine Reactor Plant in Small Electric Grids

    International Nuclear Information System (INIS)

    Khlopkin, N.; Makarov, V.; Pologikh, B.

    2002-01-01

    In report some aspects of the using marine nuclear reactor are considered for provision of need small non-interconnected power systems, as well as separate settlements and the mining enterprises disposed in regions with a undeveloped infrastructure. Recently for these purposes it is offered to use the nuclear small modular power plants. The required plant power for small electric grids lies within from 1 to several tens of MWe. Module can be collected and tested on machine-building plant, and then delivered in ready type to the working place on some transport, for instance, a barge. Through determined time it's possible to transport a module to the repair shop and also to the point of storage after the end of operation. Marine nuclear reactors on their powers, compactness, mass and size are ideal prototypes for creation of such modules. For instance, building at present floating power unit, intended for functioning in region of the Russian North, based on using reactor plants of nuclear icebreakers. Reliability and safety of the ship reactor are confirmed by their trouble-free operation during approximately 180 reactors-years. Unlike big stationary nuclear plant, working in base mode, power unit with marine reactor wholly capable to work in mode of the loading following. In contrast with reactor of nuclear icebreaker, advisable to increase the core lifetime and to reduce the enrichment of the uranium. This requires more uranium capacity fuel compositions and design of the core. In particular, possible transition from traditional for ship reactor of the channel core to cassette design. Other directions of evolution of the ship reactors, not touching the basic constructive decisions verified by practice, but promoting development of properties of self-security of plant are possible. Among such directions is reduction volumetric power density of a core. (author)

  16. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  17. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

    Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  18. Development of long-life neutron detectors for the prototype heavy water reactor 'Fugen'

    International Nuclear Information System (INIS)

    Ohteru, Shigeru; Shirayama, Shimpey.

    1981-01-01

    The development of long-life neutron detectors as the flux monitors for the prototype heavy water reactor has been made. Three kinds of neutron monitors, namely start-up monitor (SUM), power up monitor (PUM) and local power monitor (LPM), are provided. The LPM consists of 4 ion chamber type neutron detectors and a guide tube of power calibration monitor (PCM). This is useful for reactor control and fuel soundness monitor. The improvement of the neutron detectors was made for the operation under high neutron flux and gamma-ray heating. For the long-life operation, U-234 was mixed into U-235 for the conversion in the detectors. The ratio of U-234 to U-235 is 3 to 1. The PCM is also an ion chamber type detector with U-235. The mixing ratio of U-234 to U-235 was determined by a test with the JMTR. The characteristic performance was also investigated by the JMTR. After the completion of Fugen, various tests on the long-life detectors were performed with Fugen. It was hard to test the output linearity of the detectors with a large scale reactor. Therefore, it was tested that the operation range of the detectors is within the linear region of detector output. The voltage-current characteristics and the correlation of output current and saturation current were measured. The variation of the neutron sensitivity of the detectors with the cumulative dose was also studied. (Kato, T.)

  19. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  20. Application of the system engineering approach for reactor plants design

    International Nuclear Information System (INIS)

    Sitskiy, S. B.

    2010-01-01

    The main activities planned for to be implemented are: developing a data model of the reactor plant plus integration with the information model of the plant (3D model + P & ID); reengineering of processes, developing of electronic documents; description of the equipment for information management of the reactor plant lifecycle – according ISO15926

  1. Plant maintenance and advanced reactors issue, 2008

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal [ed.

    2009-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

  2. Sodium heat pipe module test for the SAFE-30 reactor prototype

    International Nuclear Information System (INIS)

    Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

    2001-01-01

    Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

  3. International review on safety requirements for the prototype fast breeder reactor “Monju”

    International Nuclear Information System (INIS)

    2016-01-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  4. Gas and water permeability of concrete for reactor buildings--prototype scale specimens

    International Nuclear Information System (INIS)

    Mills, R.H.

    1987-02-01

    The permeability testing was performed on four concrete cylinders, 0.25 m in diameter and 2 m long, modelling the wall-thickness of reactor containment structures on the prototype scale. Tests were performed on the cylinders before and after artificial induction of longitudinal cracks, intented to model defects developing after some period of adverse service conditions. Permeability increased greatly with the introduction of longitudinal cracks in the concrete, and was also affected by moisture content and casting direction. The influence of reinforcing steel could not be resolved within the bounds of experimental variability. Ultrasound measurements were taken on each cylinder before and after cracking, and a correlation between increased permeability and lowered Ultrasonic Pulse Velocity was observed. Ultrasonic Pulse Velocity measurements thus show promise as a means of continuous monitoring of the integrity of the concrete barrier in service

  5. International review on safety requirements for the prototype fast breeder reactor “Monju” (Translated document)

    International Nuclear Information System (INIS)

    2016-02-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  6. Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Bhaduri, A.K.; Indira, R.; Albert, S.K.; Rao, B.P.S.; Jain, S.C.; Asokkumar, S.

    2004-01-01

    Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K

  7. Core designs for new VVER reactors and operational experience of immediate prototypes

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Mokhov, V.; Ryzhov, S.

    2011-01-01

    The paper covers the recent improvements analyzed in order to implement the enhanced core performances. AES-2006 reactor core design is considered from the point of view of its application and improvement in the planned VVER-TOI project and of the possibilities of using the basic engineering solutions for the cores with spectral control. The discussion of several types of mixing grids considered in the paper involves a preliminary assessment of their efficiency and the information on their introduction into pilot operation at the VVER-1000 Units. Special attention is given to the results of the operation of immediate prototypes (TVS-2 and TVS-2M) that corroborate the reliability of the design both with regard for the core geometrical stability and fuel cladding tightness

  8. Real Time Computer for Plugging Indicator Control of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Manimaran, M.; Manoj, P.; Shanmugam, A.; Murali, N.; Satya Murty, S.A.V.

    2013-06-01

    Prototype Fast Breeder Reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Liquid sodium is used as coolant to transfer the heat produced in the reactor core to steam water circuit. Impurities present in the sodium are removed using purification circuit. Plugging indicator is a device used to measure the purity of the sodium. Versa Module Europa bus based Real Time Computer (RTC) system is used for plugging indicator control. Hot standby architecture consisting of dual redundant RTC system with switch over logic system is the configuration adopted to achieve fault tolerance. Plugging indicator can be controlled in two modes namely continuous and discontinuous mode. Software based Proportional-Integral-Derivative (PID) algorithms are developed for plugging indicator control wherein the set point changes dynamically for every scan interval of the RTC system. Set points and PID constants are kept as configurable in runtime in order to control the process in very efficient manner, which calls for reliable communication between RTC system and control station, hence TCP/IP protocol is adopted. Performance of the RTC system for plugging indicator control was thoroughly studied in the laboratory by simulating the inputs and monitored the control outputs. The control outputs were also monitored for different PID constants. Continuous and discontinuous mode plots were generated. (authors)

  9. Evolution of actinides in ThO2 blanket of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Bachchan, Abhitab; Riyas, A.; Devan, K.; Puthiyavinayagam, P.

    2015-01-01

    The third stage of India's nuclear program focuses on fissile fuel production through Th- 233 U cycle in view of the better abundance and relative merits of thorium. For early introduction of Thorium into the nuclear energy system, several R and D program has started to find the best possible route of thorium utilization. Towards this, efforts were made to assess the feasibility of Th-U cycle in a fast spectrum reactor like Prototype Fast Breeder Reactor (PFBR). The effect on core neutronic parameters and actinide evolution with the replacement of depleted UO 2 in the PFBR blanket SA with thorium oxide has been studied using 3-D diffusion code FARCOB. Study shows that by the introduction of thorium blanket, core excess reactivity is coming down by ∼ 535 pcm and core breeding ratio is slightly lower than conventional oxide blanket. The distribution of region wise power production is slightly changed. Power from radial blanket is reduced from 3% to 2% while the core-1 power is increased from 49 % to 50 %. The estimated 233 U production is 7.6, 11.5 and 14.1 kg/t with 180, 360 and 540 days of irradiation respectively. (author)

  10. A new physics design of control safety rods for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

    2008-01-01

    The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

  11. Carbon-14 in reactor plant water

    International Nuclear Information System (INIS)

    Knowles, G.K.

    1979-01-01

    The method for the analysis of 14 C in reactor plant water and various waste streams previously used at the Idaho National Engineering Laboratory has been shown to be ineffective for samples which contain organic compounds. The previous method consisted of acidification and refluxing of the sample, precipitation of the liberated CO 2 , and subsequent analysis by the liquid scintillation method. The method was simple but it did not convert all compounds containing 14 C in the sample to CO 2 . The new method, while it is based on the previous method, has been improved by employing a strong oxidant, potassium persulfate and silver nitrate, for more complete oxidation of the organics to CO 2 . The new method yields 14 C values that have typically been one to two orders of magnitude higher than the values obtained using the former method. This indicates that most of the 14 C present in the current reactor water samples being analyzed is associated with trace amounts of organics

  12. Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Jayalal, M.L.; Kumar, L. Satish; Jehadeesan, R.; Rajeswari, S.; Satya Murty, S.A.V.; Balasubramaniyan, V.; Chetal, S.C.

    2011-01-01

    Highlights: → We model design optimization of a vital reactor component using Genetic Algorithm. → Real-parameter Genetic Algorithm is used for steam condenser optimization study. → Comparison analysis done with various Genetic Algorithm related mechanisms. → The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

  13. Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Kumar, L. Satish, E-mail: satish@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Jehadeesan, R., E-mail: jeha@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Rajeswari, S., E-mail: raj@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Satya Murty, S.A.V., E-mail: satya@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Balasubramaniyan, V.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India)

    2011-10-15

    Highlights: > We model design optimization of a vital reactor component using Genetic Algorithm. > Real-parameter Genetic Algorithm is used for steam condenser optimization study. > Comparison analysis done with various Genetic Algorithm related mechanisms. > The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

  14. Dismantling and decontamination of Piver prototype vitrification plant

    International Nuclear Information System (INIS)

    Jouan, A.; Roudil, S.; Thomas, F.

    1991-01-01

    The PIVER prototype was targeted for dismantling in order to install a new pilot facility for the french continuous vitrification process. Most of the work involved the vitrification cell containing the process equipments, which had to be cleared out and thoroughly decontaminated; this implied disassembling, cutting up, conditioning and removing all the equipment installed in the cell. Remote manipulation, handling and cutting devices were used and some prior modifications were implemented in the cell environment. The dismantling procedure was conducted under a detailed programme defining the methodology for each operation. After equipment items and active zones were identified, the waste materials were removed, and several liquid decontamination operations were implemented. Removed activity, levels of irradiation in the cell and doses integrated by personnel were monitored to control progress and to adapt procedures to the conditions encountered. At the end of December 1989, the PIVER cleanup programme was at 87% complete and the total activity removed was 2.11 X 10 14 Bq (5712 Ci). The objective now is to obtain suitable working conditions in order to allow operators to enter the cell to remove items that are inaccessible or which cannot be dismantled by remote manipulators and to complete the decontamination procedure

  15. A prototype expert system 'SMART' for water chemistry control in reactor water circuits

    International Nuclear Information System (INIS)

    Rangarajan, S.; Narasimhan, S.V.

    1998-01-01

    The operational safety of a power plant depends mainly on the material compatibility of the system materials with the environment. However, for an operating plant, the material is almost fixed and hence one can improve the safety by controlling the surrounding environment. From the economy point of view, the plant availability factor as well as plant life extension (PLEX) are important considerations and these necessitate a systematic approach for continuous parametric monitoring, rapid data analysis and diagnosis for controlling the water chemistry regime. A prototype expert system 'SMART' was developed in BASIC language. The expert system consists of four modules. The DATA HANDLER module controls all the data handling functions and graphical display of the data parameters. It also generates weekly and monthly reports of the water chemistry data. The DATA INTERPRETER module compares the experimental data with the theoretically calculated values and predicts the presence of impurity ingress in the system. The CHEMISTRY EXPERT contains the knowledge base about the various sub-systems. All the water chemistry specifications are translated in the form of IF... THEN.. rules and are stored in this module. The expert system inferences with the forward chain reasoning mechanism to identify the diagnostic parameters by consulting the knowledge base and applying the appropriate rules. The ACTION EXPERT module collects all the diagnostic parameters and suggests the operator, the remedial actions/counter measures that should be taken immediately. This rule based system can be expanded to accommodate different water chemistry regimes. (author)

  16. Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  17. Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Sang June Ahn

    2016-08-01

    Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  18. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev, E-mail: baldev.dr@gmail.com

    2015-09-15

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  19. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-05-17

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and

  20. Inducer pumps for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Jackson, E.D.

    2002-01-01

    Pumps proposed for liquid metal reactor plants typically use centrifugal impellers as the rotating element and are required to maintain a relatively low speed to keep the suction specific speed low enough to operate at the available net positive suction head (HPSH) and to avoid cavitation damage. These low speeds of operation require that the pump diameter increase and/or multiple stages be used to achieve the design head. This frequently results in a large, heavy, complex pump design. In addition, the low speed results in a larger drive motor size so that the resultant penalty to the plant designer is multiplied. The heavier pump can also result in further complications as, e.g., the difficulty in maintaining the first critical speed sufficiently above the pump operating range to provide margin for rotor dynamic stability. To overcome some of these disadvantages, it was proposed the use of inducer pumps for Liquid Metal Fast Breeder Reactor (LMFBR) plants. This paper discusses some of the advantages of the inducer pump and the development history of designing and testing these pumps both in water and sodium. The inducer pump is seen to be a sound concept with a strong technology base derived from the aerospace and ship propulsion industries. The superior suction performance capability of the inducer offers significant system design advantages, primarily a smaller, lighter weight, less complex pump design with resulting saving in cost. Extensive testing of these pumps has been conducted in both sodium and water to demonstrate the long-life capability with no cavitation damage occurring in those designs based on Rockwell's current design criteria. These tests have utilized multiple inspection and measurement approaches to accurately assess and identify any potential for cavitation damage, and these approaches have all concluded that no damage is occurring. Therefore, it is concluded that inducer pumps can be safely designed for long life operation in sodium with

  1. Hydraulic experiments on the failed fuel location module of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Rajesh, K.; Kumar, S.; Padmakumar, G.; Prakash, V.; Vijayashree, R.; Rajan Babu, V.; Govinda Rajan, S.; Vaidyanathan, G.; Prabhaker, R.

    2003-01-01

    The design of Prototype Fast Breeder Reactor (PFBR) is based on sound design concepts with emphasis on intrinsic safety. The uncertainties involved in the design of various components, which are difficult to assess theoretically, are experimentally verified before design is validated. In PFBR core, the coolant (liquid sodium) enters the bottom of the fuel subassembly, passes over the fuel pins picking up the fission heat and issues in to a hot pool. If there is any breach in the fuel pins, the fission products come in direct contact with the coolant. This is undesirable and it is necessary to locate the subassembly with the failed fuel pin and to isolate it. A component called Failed Fuel Location Module (FFLM) is employed for locating the failed SA by monitoring the coolant samples coming out of each Subassembly. The coolant sample from each Subassembly is drawn by FFLM using an EM pump through sampling tube and selector valve and is monitored for the presence of delayed neutrons which is an indication of failure of the Subassembly. The pressure drop across the selector valve determines the rating of the EM Pump. The dilution of the coolant sample across the selector valve determines the effectiveness of monitoring for contamination. It is not possible to predict pressure drop across the selector valve and dilution of the coolant sample theoretically. These two parameters are determined using a hydraulic experiment on the FFLM. The experiment was carried out in conditions that simulate the reactor conditions following appropriate similarity laws. The paper discusses the details of the model, techniques of experiments and the results from the studies

  2. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  3. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Manimaran, M.; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-01-01

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored

  4. Prototype plant for nuclear process heat (PNP), reference phase

    International Nuclear Information System (INIS)

    Fladerer, R.; Schrader, L.

    1982-07-01

    The coal gasification processes using nuclear process heat being developed within the framwork of the PNP project, have the advantages of saving feed coal, improving efficiency, reducing emissions, and stabilizing energy costs. One major gasification process is the hydrogasification of coal for producing SNG or gas mixture of carbon monoxide and hydrogen; this process can also be applied in a conventional route. The first steps to develop this process were planning, construction and operation of a semi-technical pilot plant for hydrogasification of coal in a fluidized bed having an input of 100 kg C/h. Before the completion of the development phase (reference phase) describing here, several components were tested on part of which no operational experience had so far been gained; these were the newly developed devices, e.g. the inclined tube for feeding coal into the fluidized bed, and the raw gas/hydrogenation gas heat exchanger for utilizing the waste heat of the raw gas leaving the gasifier. Concept optimizing of the thoroughly tested equipment parts led to an improved operational behaviour. Between 1976 and 1980, the semi-technical pilot plant was operated for about 19,400 hours under test conditions, more than 7,400 hours of which it has worked under gasification conditions. During this time approx. 1,100 metric tons of dry brown coal and more than 13 metric tons of hard coal were gasified. The longest coherent operational phase under gasification conditions was 748 hours in which 85.4 metric tons of dry brown coal were gasified. Carbon gasification rates up to 82% and methane contents in the dry raw gas (free of N 2 ) up to 48 vol.% were obtained. A detailed evaluation of the test results provided information of the results obtained previously. For the completion of the test - primarily of long-term tests - the operation of the semi-technical pilot plant for hydrogasification of coal is to be continued up to September 1982. (orig.) [de

  5. Construction of a power plant with prototype DLN combustion turbines

    Energy Technology Data Exchange (ETDEWEB)

    Wilkinson, M.L. [CSW Energy, Dallas, TX (United States); Drummond, L.J. [Zurn NEPCO, Redmond, WA (United States)

    1996-12-31

    Design and construction of a power plant is always a difficult process and this is especially true when the main keystone, the combustion turbine engine, is being modified by the manufacturer resulting in numerous changes in the design interfaces. The development of the design and construction of the Orange Cogeneration Facility has been in parallel with major modification of the LM6000 to DLE technology (a Dry Low NO{sub x} combustion system). The Dry Low NO{sub x} Combustion System for a combustion turbine offered a means to reduce water usage, lower Zero Liquid Discharge System operating costs and reduce emissions to meet Florida Department of Environmental Protection requirements. This development was successfully accomplished by Owner, EPC contractor and Combustion Turbine Manufacturer by maintaining flexibility in the design and construction while the design interfaces and performance of the combustion turbines were being finalized.

  6. Nuclear reactor plant for production process heat

    International Nuclear Information System (INIS)

    Weber, M.

    1979-01-01

    The high temperature reactor is suitable as a heat source for carrying out endothermal chemical processes. A heat exchanger is required for separating the reactor coolant gases and the process medium. The heat of the reactor is transferred at a temperature lower than the process temperature to a secondary gas and is compressed to give the required temperature. The compression energy is obtained from the same reactor. (RW) [de

  7. Basic lay-out, arrangement and design criteria of heat components of the ''nuclear coal gasification prototype plant (PNP)''

    International Nuclear Information System (INIS)

    Pruschek, R.

    1980-01-01

    Since 1975, the companies Bergbau-Forschung GmbH, GHT Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH und Rheinische Braunkohlenwerke AG are working jointly on the Project ''Prototype Plant Nuclear Process Heat (PNP)'', with promotion of the ''Bundesminister fuer Forschung und Technologie'' and of the ''Minister fuer Wirtschaft, Mittelstand und Verkehr des Landes Nordrhein-Westfalen''. The objectives of the project are the development of a high-temperature reactor, with a core outlet temperature of 950 0 C, suitable for various process heat applications, and the development and testing of the appropriate coal gasification technology. The applied gasifications methods comprise endothermal and exothermal reactions. Therefore, various heat transfer components are to be developed. In the context of this Specialists Meeting, only those components will be discussed by which heat is transferred from primary helium to secondary helium or from helium to the working or process fluid

  8. The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

    International Nuclear Information System (INIS)

    Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

    1987-01-01

    The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

  9. Safety evaluation for the prototype Fast Breeder Reactor MONJU as a Japanese TSO

    International Nuclear Information System (INIS)

    Endo, Hiroshi

    2010-01-01

    In the safety field of fast breeder reactors (FBRs), JNES is conducting an evaluation work of the safety regulation by Nuclear and Industry Safety Agency (NISA) for the re-start of a prototype FBR MONJU. MONJU has been stopped over 14 years since 1995 due to a sodium leakage accident at a secondary heat transport system, and is now reached to the criticality on 8th of May, 2010. JNES is supporting the safety regulation work conducted by NISA based on the following activities: i) Support of the technical evaluation of the application for the establishment license prepared by Japan Atomic Energy Agency (JAEA), ii) Support of the description of the safety review report by NISA based on independent safety analyses for the major accident events such as unprotected loss-of-flow (ULOF) by employing the latest findings on the study of core disruptive accidents (CDAs) independently conducted by JNES, iii) Support of the risk-informed-regulation (RIR) such as an accident management (AM) review, iv), and Consideration on the safety regulation policy from the points of severe accidents and source-term behaviors including the cesium (Cs). The objective of this paper is to introduce the major activities of JNES in the safety domain of MONJU regulations. (author)

  10. Pulsed Nd-YAG laser welding of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Suresh Varma, P.V.; Gupta, Amit; Amit, K.; Bhatt, R.B.; Afzal, Mohd.; Panakkal, J.P.; Kamath, H.S.

    2009-02-01

    End plug welding of Prototype Fast Breeder Reactor (PFBR) fuel elements involves welding of fully Austenitic Stainless Steel (ASS) of grade D9 clad tube with 316M end plug. Pulsed Gas Tungsten Arc Welding (GTAW) is being used for the production of PFBR fuel elements at Advanced Fuel Fabrication Facility (AFFF). GTAW is an established process for end plug welding and hence adopted by many countries. GTAW has got certain limitations like heat input, arc gap sensitivity and certain sporadic defects like tungsten inclusion. Experiments have been carried out at AFFF to use Laser Beam Welding (LBW) technique as LBW offers a number of advantages over the former process. This report mainly deals with the optimization of laser parameters for welding of PFBR fuel elements. To facilitate pulsed Nd-YAG laser spot welding, parameters like peak power, pulse duration, pulse energy, frequency and defocusing of laser beam on to the work piece have been optimized. On the basis of penetration requirement laser welding parameters have been optimized. (author)

  11. Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

    International Nuclear Information System (INIS)

    Swannack, D.L.

    1982-01-01

    This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

  12. Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

    International Nuclear Information System (INIS)

    1989-09-01

    The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

  13. Gamma irradiation plants using reactor fuel elements

    International Nuclear Information System (INIS)

    Suckow, W.

    1976-11-01

    Recent irradiation plants utilizing fuel elements are described. Criteria for optimizing such plants, evaluation of the plants realized so far, and applications for the facilities are discussed. (author)

  14. Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

    International Nuclear Information System (INIS)

    Veshnyakov, K.B.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

    2000-01-01

    Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

  15. Automated Work Packages Prototype: Initial Design, Development, and Evaluation. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna Helene [Idaho National Lab. (INL), Idaho Falls, ID (United States); Al Rashdan, Ahmad [Idaho National Lab. (INL), Idaho Falls, ID (United States); Le Blanc, Katya Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bly, Aaron Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-07-01

    The goal of the Automated Work Packages (AWP) project is to demonstrate how to enhance work quality, cost management, and nuclear safety through the use of advanced technology. The work described in this report is part of the digital architecture for a highly automated plant project of the technical program plan for advanced instrumentation, information, and control (II&C) systems technologies. This report addresses the DOE Milestone M2LW-15IN0603112: Describe the outcomes of field evaluations/demonstrations of the AWP prototype system and plant surveillance and communication framework requirements at host utilities. A brief background to the need for AWP research is provided, then two human factors field evaluation studies are described. These studies focus on the user experience of conducting a task (in this case a preventive maintenance and a surveillance test) while using an AWP system. The remaining part of the report describes an II&C effort to provide real time status updates to the technician by wireless transfer of equipment indications and a dynamic user interface.

  16. Automated Work Packages Prototype: Initial Design, Development, and Evaluation. Light Water Reactor Sustainability Program

    International Nuclear Information System (INIS)

    Oxstrand, Johanna Helene; Al Rashdan, Ahmad; Le Blanc, Katya Lee; Bly, Aaron Douglas; Agarwal, Vivek

    2015-01-01

    The goal of the Automated Work Packages (AWP) project is to demonstrate how to enhance work quality, cost management, and nuclear safety through the use of advanced technology. The work described in this report is part of the digital architecture for a highly automated plant project of the technical program plan for advanced instrumentation, information, and control (II&C) systems technologies. This report addresses the DOE Milestone M2LW-15IN0603112: Describe the outcomes of field evaluations/demonstrations of the AWP prototype system and plant surveillance and communication framework requirements at host utilities. A brief background to the need for AWP research is provided, then two human factors field evaluation studies are described. These studies focus on the user experience of conducting a task (in this case a preventive maintenance and a surveillance test) while using an AWP system. The remaining part of the report describes an II&C effort to provide real time status updates to the technician by wireless transfer of equipment indications and a dynamic user interface.

  17. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

  18. Main gas circulator for VG-400 reactor plant

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kostin, V.I.; Novinskij, E.G.; Kuropatov, A.I.; Protsenko, A.N.; Smirnov, V.P.; Stolyarevskij, A.Ya.

    1988-01-01

    Principle parameters and operating conditions of the main gas circulator (MGC) in VG-400 reactor plant are presented. Brief MGC design description and experimental work scope are given. (author). 4 refs, 4 figs, 1 tab

  19. Inspection of the UO2 special fuel for the prototype heavy water reactor 'FUGEN'

    International Nuclear Information System (INIS)

    Miura, Makoto; Ohmori, Takuro; Yoshino, Hiroyuki; Matsui, Hiromasa; Hirosawa, Naonori

    1979-01-01

    UO 2 special fuel assemblies are the fuel for material irradiation incorporating irradiation specimens, for the prototype heavy water reactor ''FUGEN''. In order to monitor the behavior of the pressure tube material irradiated with neutrons for long time, monitoring specimens were equipped in the core. This special fuel was fabricated by the Nuclear Fuel Industries, Ltd. (NFI), and the fuel cladding tubes, the capsule guide tubes and the capsule tubes were furnished by PNC. The irradiation specimens were prepared by PNC, and incorporated into the assemblies by NFI. The inspection by PNC on the special fuel assemblies was conducted following the inspection by the maker, which was made on UO 2 pellets, fuel element and assembly parts except cladding tubes, after completing the fabrication. The specifications of the special fuel, especially for the outer and inner layer pellets, the outer and inner layer fuel elements and the fuel assemblies, are presented. The flow sheet for the inspection process and surveillance test of special fuel assemblies is illustrated. The inspection items, the materials and the quantity inspection are tabulated for the fuel elements, the fuel assemblies and the irradiation capsules, respectively. The structure of a special type fuel assembly is shown. For each inspection, the inspection methods and items and the results are explained. As for the results of inspection of the special fuel, the UO 2 pellets, fuel element parts, fuel elements, fuel assembly parts, fuel assemblies, capsules and irradiation specimens were in accordance with the specifications. Regarding the situation of the quality control in the processes, check was made with many documents, and it was recognized that the quality control was performed in the quality assurance program. (Nakai, Y.)

  20. Fault tolerant distributed real time computer systems for I and C of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2014-03-15

    Highlights: • Architecture of distributed real time computer system (DRTCS) used in I and C of PFBR is explained. • Fault tolerant (hot standby) architecture, fault detection and switch over are detailed. • Scaled down model was used to study functional and performance requirements of DRTCS. • Quality of service parameters for scaled down model was critically studied. - Abstract: Prototype fast breeder reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Three-tier architecture is adopted for instrumentation and control (I and C) of PFBR wherein bottom tier consists of real time computer (RTC) systems, middle tier consists of process computers and top tier constitutes of display stations. These RTC systems are geographically distributed and networked together with process computers and display stations. Hot standby architecture comprising of dual redundant RTC systems with switch over logic system is deployed in order to achieve fault tolerance. Fault tolerant dual redundant network connectivity is provided in each RTC system and TCP/IP protocol is selected for network communication. In order to assess the performance of distributed RTC systems, scaled down model was developed with 9 representative systems and nearly 15% of I and C signals of PFBR were connected and monitored. Functional and performance testing were carried out for each RTC system and the fault tolerant characteristics were studied by creating various faults into the system and observed the performance. Various quality of service parameters like connection establishment delay, priority parameter, transit delay, throughput, residual error ratio, etc., are critically studied for the network.

  1. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

  2. Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

    International Nuclear Information System (INIS)

    Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

    1977-01-01

    GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

  3. Design and fabrication of fuel for the prototype heavy water reactor Fugen

    International Nuclear Information System (INIS)

    Hasumi, Takashi; Yamanaka, Ryozi; Osawa, Masahide; Asami, Tomohiro; Kaziyama, Takashi

    1983-01-01

    For the advanced thermal reactor Fugen, 224 fuel assemblies were charged as the initial charge fuel, of which 96 were uranium-plutonium mixed oxide fuel, and 128 were uranium dioxide fuel. Since the full scale operation was started in March, 1979, fuel exchange was carried out five times, and 240 fuel assemblies were taken out, but fuel breaking was never found, and the fuel for Fugen has shown good result. For 16 mixed oxide fuel assemblies for the third exchange and thereafter, the domestically produced plutonium extracted in the Tokai reprocessing plant has been used, and for 15 UO 2 fuel assemblies for the fifth exchange, the enriched uranium produced in the Ningyo Pass plant was used. These fuels are burning in the core without causing trouble. The course of the development of the fuel is described as follows: trial manufacture, evaluation test outside the core, heat transferring flow characteristic test, irradiation test, design of fuel elements and fuel assemblies, production of fuel and quality assurance, and results of production and use. (Kako, I.)

  4. Analysis of reactor trips involving balance-of-plant failures

    International Nuclear Information System (INIS)

    Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

    1986-01-01

    The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

  5. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    Lerner, Ana Maria; Madariaga, Marcelo

    2002-01-01

    The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

  6. A prototype for actinide alpha monitoring in liquid effluents of reprocessing plants

    International Nuclear Information System (INIS)

    Bardone, G.; Mattia, B.; Durante, R.; Frazzoli, F.V.

    1983-01-01

    The report deals with the design criteria of prototype measuring device, based on the alpha spectrometry, aimed to the determination of actinides solutions in reprocessing plants. The described instrument is considered as the result of a preliminary stage of development. Taking into account the experimental results obtained with Pu bearing solutions the performances achievable are evaluated; in particular, it turns out that the minimum detectable activity is about 10 -5 Ci/l

  7. The manual of a computer software 'FBR Plant Planning Design Prototype System'

    International Nuclear Information System (INIS)

    2003-10-01

    This is a manual of a computer software 'FBR Plant Planning Design Prototype System', which enables users to conduct case studies of deviated FBR design concepts based on 'MONJU'. The calculations simply proceed as the user clicks displayed buttons, therefore step-by-step explanation is supposed not be necessary. The following pages introduce only particular features of this software, i.e, each interactive screens, functions of buttons and consequences after clicks, and the quitting procedure. (author)

  8. Reactor shutdown: nuclear power plant performance

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The article essentially looks at the performance of nine of Sweden's nuclear reactors. A table lists the percentage of time for the first three quarters of 1981 that the reactors were operating, and the number of hours out of service for planned or other reasons. In particular, one station - Ringhals 3 - was out of action because of a damaged tube in the associated steam generator. The same fault occurred with another reactor - Ringhals 4 - before this was brought into service. The reasons for the failure and its importance are briefly discussed. (G.P.)

  9. Fast reactor system factors affecting reprocessing plant design

    International Nuclear Information System (INIS)

    Allardice, R.H.; Pugh, O.

    1982-01-01

    The introduction of a commercial fast reactor electricity generating system is very dependent on the availability of an efficient nuclear fuel cycle. Selection of fuel element constructional materials, the fuel element design approach and the reactor operation have a significant influence on the technical feasibility and efficiency of the reprocessing and waste management plants. Therefore the fast reactor processing plant requires liaison between many design teams -reactor, fuel design, reprocessing and waste management -often with different disciplines and conflicting objectives if taken in isolation and an optimised approach to determining several key parameters. A number of these parameters are identified and the design approach discussed in the context of the reprocessing plant. Radiological safety and its impact on design is also briefly discussed. (author)

  10. Plant to reduce reactor scrams and ASSET related analysis

    Energy Technology Data Exchange (ETDEWEB)

    Piirto, A [Teollisuuden Voima Oy (Finland)

    1997-10-01

    The report of events at the plant follows established rules. Basically, three categories of reports exist: first, the reactor scram report, secondly, the operational disturbance report and thirdly, the special report. The last named category covers events defined by authorities. It concentrates on safety related events, for example on failures to follow the requirements stipulated in the plant`s Technical Specifications. In general, special events are nuclear safety on the plant, the safety of the plant personnel or overall, the radiation safety in the plant`s vicinity. In the TVO`s experience feedback activity the greatest emphasis is put on events at the TVO plant. The events on the same type of plants come second. Due to limited resources, somewhat less attention is paid to events on the other types of plants. However, the experience feedback should become wider in practice so that it would be a part of everyday life in nuclear power plant operation.

  11. 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

  12. Clinch River Breeder Reactor Plant Project: construction schedule

    International Nuclear Information System (INIS)

    Purcell, W.J.; Martin, E.M.; Shivley, J.M.

    1982-01-01

    The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

  13. Interactive nuclear plant analyzer for VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Horak, W.; Kennett, R.

    1992-05-01

    This document discusses an interactive nuclear plant analyzer (NPA) which has been developed for a VVER-440, Model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator

  14. Power-up of Fugen reactor and development of demonstration plant

    International Nuclear Information System (INIS)

    Sawai, Sadamu; Akebi, Michio; Yazaki, Akira.

    1979-06-01

    The Fugen Nuclear Power Station is the 165 MWe prototype plant characterized by heavy water-moderated, boiling light water-cooled, pressure tube type, and was developed by the Power Reactor and Nuclear Fuel Development Corporation, Japan. The plant went into commercial operation on March 20, 1979, in Tsuruga, Fukui Prefecture. Some delay in the overall schedule occurred due to the shortage of cement caused by the oil crisis, more stringent regulations as the result of stress corrosion cracking experienced in BWRs, and design modifications. All functional tests, the final check-up of the whole plant, and remaining modifying works had been completed by March 10, 1978. The minimum criticality was achieved with 22 mixed oxide fuel assemblies on March 20, 1978. Thereafter, the tests on reactor physics, plant dynamics, the performances of components and systems, and radiation and water chemistry have been carried out. 5 MWe was sent to grid system for the first time on July 29, 1978. The commercial operation of the plant was licenced by the Government on March 30, 1979. The conceptual design of the 600 MWe demonstration plant was finished in early 1979, and the detailed design is to be carried out in 1979 and 1980. The main design principle was incorporated in the conceptual design, but some modifications are to be made to reduce the power cost and to facilitate the easy maintenance. (Kako, I.)

  15. Prototype of an expert system based nuclear power plant information systems

    International Nuclear Information System (INIS)

    Vegh, J.; Bodnar, M.; Buerger, L.; Tanyi, M.; Sefesik, F.

    1994-01-01

    The components and functioning of the GPCS information system applicable for intelligent process monitoring and alarm generation in a WWER-440 type nuclear power plant are described. The prototype system has been developed by using the G2 expert system, plant measurements were simulated by a WWER-440 compact simulator and by archive replay sessions performed by the VERONA-u core monitoring system. The GPCS contains an object oriented description of the basic subsystems of the plant and concentrates on the fast evaluation/displaying of measurements and alarms. The high-level information reflecting actual plant safety status is synthesized from primary measured data, by forming global alarms and by evaluating logical diagrams. (author). 10 refs, 4 figs

  16. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  17. A decision support system for maintenance management of a boiling-water reactor power plant

    International Nuclear Information System (INIS)

    Shen, J.H.; Ray, A.; Levin, S.

    1996-01-01

    This article reports the concept and development of a prototype expert system to serve as a decision support tool for maintenance of boiling-water reactor (BWR) nuclear power plants. The code of the expert system makes use of the database derived from the two BWR units operated by the Pennsylvania Power and Light Company in Berwick, Pennsylvania. The operations and maintenance information from a large number of plant equipment and sub-systems that must be available for emergency conditions and in the event of an accident is stored in the database of the expert system. The ultimate goal of this decision support tool is to identify the relevant Technical Specifications and management rules for shutting down any one of the plant sub-systems or removing a component from service to support maintenance. 6 refs., 7 figs

  18. Balance of plant improvements for future reactor projects

    International Nuclear Information System (INIS)

    Hollingshaus, H.

    1987-01-01

    Many studies have shown that improvements in portions of the plant other than the reactor systems can yield large cost savings during the design, construction, and operation of future reactor power plants. This portion is defined as the Balance of Plant which includes virtually everything except the equipment furnished by the Nuclear Steam Supply System manufacturer. It normally includes the erection of the entire plant including the NSSS. Cost of BOP equipment, engineering and construction work is therefore most of the cost of the plant. Improvements in the BOP have been identified that will substantially reduce nuclear plant cost and construction time while at the same time increasing availability and operability and improving safety. Improvements achieved through standardizatoin, simplification, three-dimensional (3D) computer-aided design, modular construction, innovative construction techniques, and applications for Artificial Intelligence Systems are described. (author)

  19. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  20. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Journeau, Ch.

    2008-01-01

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  1. Adapting a reactor safety assessment system for specific plants

    International Nuclear Information System (INIS)

    Ballard, T.L.; Cordes, G.A.

    1991-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

  2. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

    1978-12-01

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  3. Plant to reduce reactor scrams and ASSET related analysis

    International Nuclear Information System (INIS)

    Piirto, A.

    1997-01-01

    The report of events at the plant follows established rules. Basically, three categories of reports exist: first, the reactor scram report, secondly, the operational disturbance report and thirdly, the special report. The last named category covers events defined by authorities. It concentrates on safety related events, for example on failures to follow the requirements stipulated in the plant's Technical Specifications. In general, special events are nuclear safety on the plant, the safety of the plant personnel or overall, the radiation safety in the plant's vicinity. In the TVO's experience feedback activity the greatest emphasis is put on events at the TVO plant. The events on the same type of plants come second. Due to limited resources, somewhat less attention is paid to events on the other types of plants. However, the experience feedback should become wider in practice so that it would be a part of everyday life in nuclear power plant operation

  4. Liquid metal reactor/Pressurized water reactor plant comparison study

    International Nuclear Information System (INIS)

    Halverson, T.G.

    1986-01-01

    The selection between alternative electric power generating technologies is mainly based on their overall economics. Capital costs account for over 60% of the total busbar cost of nuclear plants. Estimates reported in the literature have shown capital cost ratios of LMRs to PWRs ranging from less than 1 to as high as 1.8. To reduce this range of uncertainty, the study selected a method for cataloging plant hardware and then performed comparisons using engineering judgment as to the anticipated and reasonable cost differences. The paper summarizes the resulting one-on-one comparisons of components, systems, and buildings and identifies the LMR-PWR similarities and differences which influence costs. The study leads to the conclusion that the capital cost of the most up-to-date large LMR design would be very close to that of the latest PWRs

  5. Training nuclear power plant personnel on SR-O reactor

    International Nuclear Information System (INIS)

    Cerny, K.; Boucek, F.; Kveton, M.; Prokopec, Z.; Fleischhans, J.

    1983-01-01

    The SR-O reactor is an experimental pool-type reactor with a maximum output of 1 MW and maximum thermal neutron flux density of 5.3x10 13 m -2 s -1 . The reactor is described in detail and its specifications are given. The protection and control systems of the reactor permit both manual and automatic operation. The reactor is used for training courses for nuclear power plant operators and for post-graduate study courses for other specialists. Intensive courses for 4 to 6 persons take 15 to 20 days. The course is adjusted to the results of introductory theoretical tests. An optimal teaching method has been developed based on the flowchart algorithmic method, dividing activities into operations (manipulations with controls, issuing commands, making records, etc.) and decision making (information reception and processing). (M.D.)

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

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

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

  7. Tritium instrumentation for a fusion reactor power plant

    International Nuclear Information System (INIS)

    Shank, K.E.; Easterly, C.E.

    1976-09-01

    A review of tritium instrumentation is presented. This includes a discussion of currently available in-plant instrumentation and methods required for sampling stacks, monitoring process streams and reactor coolants, analyzing occupational work areas for air and surface contamination, and personnel monitoring. Off-site instrumentation and collection techniques are also presented. Conclusions are made concerning the adequacy of existing instrumentation in relation to the monitoring needs for fusion reactors

  8. High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Venkatraman, B.; Sethi, V.K.; Jayakumar, T.; Raj, B.

    1995-01-01

    In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

  9. Analysis of reactor trips originating in balance of plant systems

    International Nuclear Information System (INIS)

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W.

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs

  10. Potential of light water reactors for future nuclear power plants

    International Nuclear Information System (INIS)

    Gueldner, R.

    2003-01-01

    Energy consumption worldwide is going to increase further in the next few decades. Reliable supplies of electricity can be achieved only by centralized power plant structures. In this scenario, nuclear power plants are going to play a leading role as reliable and competitive plants, also under deregulated market conditions. Today, light water reactors have achieved a leading position, both technically and economically, contributing 85% to worldwide electricity generation in nuclear plants. They will continue to be a proven technology in power generation. In many countries, activities therefore are concentrated on extending the service life of plants beyond a period of forty years. New nuclear generating capacities are expected to be created and added from the end of this decade onward. Most of this capacity will be in light water reactors. The concepts of third-generation reactors will meet all economic and technical safety requirements of the 21st century and will offer considerable potential for further development. Probably some thirty years from now, fourth-generation nuclear power plants will be ready for commercial application. These plants will penetrate especially new sectors of the energy markets. Public acceptance of new nuclear power plants is not a matter of reactor lines, provided that safety requirements are met. The important issue is the management of radioactive waste. The construction of new nuclear power plants in Western Europe and North America mainly hinges on the ability to explain to the public that there is a need for new plants and that nuclear power is fundamental to assuring sustainable development. (orig.)

  11. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

  12. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies

  13. Development step toward fusion power plant and role of experimental reactor ITER

    International Nuclear Information System (INIS)

    Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

    2005-01-01

    The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

  14. Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

  15. Interactive nuclear plant analyzer for the VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Kennett, R.

    1993-01-01

    An interactive nuclear plant analyzer (NPA) has been developed for a VVER-440 model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator. Results of the interactive calculation can be through the user-defined, digital display of various plant parameters and through color changes that reflect changes in primary system fluid temperatures, fuel and clad temperatures, and the temperatures of other metal structures. In addition, changes in the status of various components and system can be initiated and/or displayed both numerically and graphically on the mask

  16. Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials

    International Nuclear Information System (INIS)

    Binder, J.L.; McUmber, L.M.; Spencer, B.W.

    1993-01-01

    Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1B and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests

  17. 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

  18. Plant Control Concept for the Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Kim, S. O.

    2010-12-01

    A power plant is designed for incorporation into a utility's grid system and follows the load demand through the steam generator, intermediate heat exchanger(IHX), from the nuclear core. During the load-following transients, various plant parameters must be controlled to protect the reactor core and other components in the plant. The purpose of this report is to review design considerations to establish SFR plant control and to design plant control concepts. The governing equations and solution procedure of the computer code to calculate plant temperature conditions during the part-load operation was reviewed and 4 types of plant operation concepts were designed, and the results of the calculations were compared

  19. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    1982-01-01

    The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

  20. Fabrication of a pressurized water reactor fuel element prototype with Zy-control rod guide tubes

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1978-10-01

    A prototype fuel assembly with zircaloy guide was fabricated by mass production methods. The fastening of the Inconel spacer grids to the guide tubes and the transition joint for fixing the tubes to the stainless stell upper end-fitting of the assembly were investigated. Tools and welding devices were developed for the construction of the skeleton. (orig.) [de

  1. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  2. L-Reactor operation, Savannah River Plant: environmental assessment

    International Nuclear Information System (INIS)

    1982-08-01

    The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

  3. ELMO Bumpy Torus Reactor and power plant: conceptual design study

    International Nuclear Information System (INIS)

    Bathke, C.G.; Dudziak, D.J.; Krakowski, R.A.

    1981-08-01

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based on these results, the direction for future improvement on an already attractive reactor design is identified

  4. 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.)

  5. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

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

  6. A comparison of prototype compound parabolic collector-reactors (CPC) on the road to SOLARDETOX technology.

    Science.gov (United States)

    Funken, K H; Sattler, C; Milow, B; De Oliveira, L; Blanco, J; Fernández, P; Malato, S; Brunott, M; Dischinge, N; Tratzky, S; Musci, M; de Oliveira, J C

    2001-01-01

    Solar photocatalytic detoxification of non-biodegradable chlorinated hydrocarbon solvents (NBCS) is carried out in different concentrating and non concentrating devices using TiO2 as a photocatalyst fixed on the inner surface of the reaction tubes or as a slurry catalyst which has to be removed from the treated water. The reaction is most effective using 200 mg/l of TiO2 as a slurry in a non concentrating CPC reactor. The concentrating parabolic trough reactor has a poor activity because of its minor irradiated reactor surface. Catalyst coated glass tubes are less efficient then the used slurry catalyst. Their advantage is that no catalyst has not to be removed from the treated water and there is no loss of activity during treatment. Yet their physical stability is not sufficient to be competitive to the slurry catalyst. Nevertheless the degradation results are very promising and will possibly lead to commercial applications of this technology.

  7. Steam plant for pressurized water reactors

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

  8. NetCDF based data archiving system applied to ITER Fast Plant System Control prototype

    International Nuclear Information System (INIS)

    Castro, R.; Vega, J.; Ruiz, M.; De Arcas, G.; Barrera, E.; López, J.M.; Sanz, D.; Gonçalves, B.; Santos, B.; Utzel, N.; Makijarvi, P.

    2012-01-01

    Highlights: ► Implementation of a data archiving solution for a Fast Plant System Controller (FPSC) for ITER CODAC. ► Data archiving solution based on scientific NetCDF-4 file format and Lustre storage clustering. ► EPICS control based solution. ► Tests results and detailed analysis of using NetCDF-4 and clustering technologies on fast acquisition data archiving. - Abstract: EURATOM/CIEMAT and Technical University of Madrid (UPM) have been involved in the development of a FPSC (Fast Plant System Control) prototype for ITER, based on PXIe (PCI eXtensions for Instrumentation). One of the main focuses of this project has been data acquisition and all the related issues, including scientific data archiving. Additionally, a new data archiving solution has been developed to demonstrate the obtainable performances and possible bottlenecks of scientific data archiving in Fast Plant System Control. The presented system implements a fault tolerant architecture over a GEthernet network where FPSC data are reliably archived on remote, while remaining accessible to be redistributed, within the duration of a pulse. The storing service is supported by a clustering solution to guaranty scalability, so that FPSC management and configuration may be simplified, and a unique view of all archived data provided. All the involved components have been integrated under EPICS (Experimental Physics and Industrial Control System), implementing in each case the necessary extensions, state machines and configuration process variables. The prototyped solution is based on the NetCDF-4 (Network Common Data Format) file format in order to incorporate important features, such as scientific data models support, huge size files management, platform independent codification, or single-writer/multiple-readers concurrency. In this contribution, a complete description of the above mentioned solution is presented, together with the most relevant results of the tests performed, while focusing in the

  9. 1980 nuclear plant survey: no reactors sold; more cancellations

    International Nuclear Information System (INIS)

    Friedlander, G.D.

    1980-01-01

    No sales were reported in 1979 by any of the big four reactor suppliers. Three cancellations were reported and construction was suspended on the Jersey Central Power and Light Co.'s Forked River unit. Since last year's survey, the commercial operation dates of about 80 units have been postponed from one year to indefinitely, and nuclear commitments are down from last year's 195 units to 193 units. Presently, there are 72 plants on line, with a capacity of more than 53,000 MW. A resumption of new reactor orders is expected in either late 1980 or early 1981

  10. Multifrequency tests in the EBR-II reactor plant

    International Nuclear Information System (INIS)

    Feldman, E.E.; Mohr, D.; Gross, K.C.

    1989-01-01

    A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

  11. Start-up analysis of INET-5 MW district heating prototype reactor

    International Nuclear Information System (INIS)

    Li Tianshu

    1991-09-01

    The main features and thermohydraulic design parameters of the INET-5 MW reactor (INET: Institute of Nuclear Technology of Tsinghua University, Beijing) are presented. The start-up process and the effect of thermohydraulic instability on start-up process have been analyzed. The main obstacle of start-up process of INET-5 MW reactor is to pass the instability region from 1 atm to normal operation condition. For avoiding instability, the start-up process should be divided into two steps. The results of three different start-up proposals calculated by DACOL code are given and compared. The possibility of instabilities for each proposal has been checked. The checked results show that there is no instability during start-up of the three proposals. So, it is supposed that the INET-5 MW reactor can safely and stably reach the operation conditions. Finally, some conclusions about the effect of instability on start-up in boiling mode of INET-5MW reactor are given

  12. Prototype plant for nuclear process heat (PNP) - operation of the pilot plant for hydrogasification of coal

    International Nuclear Information System (INIS)

    Bruengel, N.; Dehms, G.; Fiedler, P.; Gerigk, H.P.; Ruddeck, W.; Schrader, L.; Schumacher, H.J.

    1988-04-01

    The Rheinische Braunkohlenwerke AG developed the process of hydrogasification of coal in a fluidized bed for generation of SNG. On basis of test results obtained in a semi-technical pilot plant of a through-put of 250 kg/h dried coal a large pilot plant was erected processing 10 t/h dried brown coal. This plant was on stream for about 14700 h, of which about 7800 h were with gasifier operation; during this time about 38000 t of dried brown coal of the Rhenish district were processed containing 4 to 25% of ash. At pressures of 60 to 120 bar and temperatures of 800 to 935 0 C carbon conversion rates up to 81 percent and methane amounts of 5000 m 3 (STP)/h were reached. The decisive parameter for methane generation was the hydrogen/coal-ratio. Even at high moisture contents, usually diminishing the methane yield from the coal essentially, by high hydrogen/coal-ratios high methane yields could be obtained. The gasifier itself caused no troubles during the total time operation. Difficulties with the original design of the residual char cooler could be overcome by change-over from water injection to liquid carbon dioxide. The design of the heat recovery system proved well. Alltogether so the size increasement of the gasifier from the semi-technical to the large pilot plant as well as the harmonization of gas generation and gas refining was proved. (orig.) With 20 refs., 20 tabs., 81 figs [de

  13. Fabrication and quality assurance of some important components and sub-assemblies for Prototype Fast Breeder Reactor (PFBR) project

    International Nuclear Information System (INIS)

    Dutta, N.G.; More, S.S.

    2010-01-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500 MWe prototype fast breeder reactor (PFBR) at Kalpakkam, Chennai. In this very important and prestigious national programmed M/s Kay Bouvet Engg. Pvt. Ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies. M/s KBEPL is engaged in manufacturing, quality assurance and supply of many subassemblies of PFBR like under water trolley (UWT), shielding door, container and container storage rack (CSR), vessel in fuel transfer cell (FTC), personnel air lock (PAL), emergency air lock (EAL) and material air lock (MAL), absorber rod drive mechanism (ARDM) flask assembly and carriage in MAL etc. Two partition doors and four nos. of embedded parts (SS 304L) have already been supplied to Bhavini. The paper deals with manufacturing and Q.A. activities being carried out for supply of these important assemblies to PFBR projects. (author)

  14. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

    2017-06-15

    Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

  15. Examples of CEA managements of spent fuels from a prototype power reactor (PHENIX) and from commercial power reactors after post irradiation examinations

    International Nuclear Information System (INIS)

    Guay, P.

    1988-01-01

    CEA gained a good experience in the management of spent fuels from its research or power prototype reactors and of the fuel samples for post irradiation examinations. The solution for these products is the reprocessing. The delay to apply that solution is bound to the disponibility of the reprocessing facilities, and in several cases induce a delayed reprocessing. Only particular and limited fuels are planned to be sent in a definitive storage. The definitive storage is choosen only for a few fuels essentially requiring important modifications of the dissolution process. The treatments and operations on the spent fuels must be carried out following the French safety rules. Long and detailed flowsheet studies are therefore necessary before the setting up of the operations. Generally the cost of the management of limited quantities of fuels, as it is the case here, is high. The flowsheets are established in taking into account, as far as possible, the use of existing facilities, procedures, transport casks

  16. Reactor plant construction productivity, why so different

    International Nuclear Information System (INIS)

    Palmeter, S.B.

    1976-01-01

    The manual labor component (manhours per kw) required to construct a nuclear power plant has increased radically since the advent of the fixed price turnkey projects of the late 1960's and early 1970's. Utilities and their architect-engineers have been, for the past several years, evaluating and diagnosing possible reasons for the increase and, in particular, the wide variation in labor manhours per kw among plants built in the same time frame. Since construction labor can amount to as much as 35--40% of direct capital cost, ways and means must be found to arrest this manhour escalation. One important way is by improving productivity. Some of the manhour increase is beyond an owner's control, e.g. NRC regulatory and other federal and state requirements adding to the scope of work. Several areas where there is potential for productivity improvement are identified as follows: (1) Revise contract strategy and bid work on a fixed price basis. This can be done by utilizing bid packages where the scope of work is clearly identified and based on well defined plans and specifications. (2) Upgrade the quality of construction management and remove first line supervision from union control. Use periodic work sampling to pinpoint causes and cure for poor productivity. (3) Reduce design complexity and improve constructibility by means of innovative design and material utilization--models help. (4) Improve labor productivity by restoring management rights in collective bargaining agreements. If this is not possible, go open shop or owner build with your own work force

  17. Practice of knowledge management in Prototype and Plant Development Center (PDC)

    International Nuclear Information System (INIS)

    Mohamad Safuan Sulaiman; Rapieh Aminuddin; Rosli Darmawan; Mohd Ashhar Khalid

    2007-01-01

    As reflecting the evolvement and movement of world economy direction, Malaysia move one step a head towards knowledge based economy (K-Economy). The movement indirectly changes the Malaysian Nuclear Agency (Nuclear Malaysia) environment to contribute to the K-Economy in the field of science and technology. Therefore, the practice of knowledge management is slowly introduced to the Nuclear Malaysia community to support the K-Economy. This paper describes the detail of the practice of knowledge management at macro and micro level in an organization. The Prototype and Plant Development Center(PDC) under the Technical Support Division, Technical Service Program has been chosen to be the case study in implementing the practice of knowledge management in Nuclear Malaysia. The main objective of this paper is to introduce the right practice of Knowledge management in an organization and PDC as among the first case for this purpose. (Author)

  18. Radiological considerations in the design of Reprocessing Uranium Plant (RUP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

    2012-01-01

    A Fast Reactor Fuel Cycle Facility (FRFCF) being planned at Indira Gandhi Centre for Atomic Research, Kalpakkam is an integrated facility with head end and back end of fuel cycle plants co-located in a single place, to meet the refuelling needs of the prototype fast breeder reactor (PFBR). Reprocessed uranium oxide plant (RUP) is one such plant in FRFCF to built to meet annual requirements of UO 2 for fabrication of fuel sub-assemblies (FSAs) and radial blanket sub-assemblies (RSAs) for PFBR. RUP receives reprocessed uranium oxide powder (U 3 O 8 ) from fast reactor fuel reprocessing plant (FRP) of FRFCF. Unlike natural uranium oxide plant, RUP has to handle reprocessed uranium oxide which is likely to have residual fission products activity in addition to traces of plutonium. As the fuel used for PFBR is recycled within these plants, formation of higher actinides in the case of plutonium and formation of higher levels of 232 U in the uranium product would be a radiological problem to be reckoned with. The paper discussed the impact of handling of multi-recycled reprocessed uranium in RUP and the radiological considerations

  19. Computer Aided Analysis and Prototype Testing of an Improved Biogas Reactor For Biomass System

    Directory of Open Access Journals (Sweden)

    Jeremy (Zheng Li

    2015-05-01

    Full Text Available The alternative fuel resources substituting for conventional fuels are required due to less availability of fuel resources than demand in the market. A large amount of crude oil and petroleum products are required to be imported in many countries over the world. Also the environmental pollution is another serious problem when use petroleum products. Biogas, with the composition of 54.5% CH4, 39.5% CO2, and 6% other elements (i.e., H2, N2, H2S, and O2, is a clear green fuel that can substitute the regular petroleum fuels to reduce the pollutant elements. Biogas can be produced by performing enriching, scrubbing, and bottling processes. The purification process can be further applied to take away the pollutants in biogas. The pure biogas process analyzed in this research is compressed to 2950 psi while being filled into gas cylinder. The daily produced biogas capacity is around 5480 ft3 and the processing efficacy is affected by surrounding environment and other factors. The design and development of this biogas system is assisted through mathematical analysis, 3D modeling, computational simulation, and prototype testing. Both computer aided analysis and prototype testing show close results which validate the feasibility of this biogas system in biomass applications.

  20. Prototype CIRCE plant-industrial demonstration of heavy-water production from a reformed hydrogen source

    Energy Technology Data Exchange (ETDEWEB)

    Spagnolo, D.A.; Boniface, H.A.; Sadhankar, R.R.; Everatt, A.E.; Miller, A.I. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Blouin, J. [Air Liquide Canada, Hamilton, Ontario (Canada)

    2002-09-01

    Heavy-water (D{sub 2}0) production has been dominated by the Girdler-Sulphide (G-S) process, which suffers several intrinsic disadvantages that lead to high production costs. Processes based on hydrogen/water exchange have become more attractive with the development of proprietary wetproofed catalysts by AECL. One process that is synergistic with industrial hydrogen production by steam methane reforming (SMR), the combined industrial reforming and catalytic exchange (CIRCE) process, offers the best prospect for commercialization. SMRs are common globally in the oil upgrading and ammonia industries. To study the CIRCE process in detail, AECL, in collaboration with Air Liquide Canada, constructed a prototype CIRCE plant (PCP) in Hamilton, ON. The plant became fully operational in 2000 July and is expected to operate to at least the late fall of 2002. To date, plant operation has confirmed the adequacy of the design and the capability of enriching deuterium to produce heavy water without compromising hydrogen production. The proprietary wetproofed catalyst has performed as expected, both in activity and in robustness. (author)

  1. Prototype CIRCE plant-industrial demonstration of heavy-water production from a reformed hydrogen source

    International Nuclear Information System (INIS)

    Spagnolo, D.A.; Boniface, H.A.; Sadhankar, R.R.; Everatt, A.E.; Miller, A.I.; Blouin, J.

    2002-09-01

    Heavy-water (D 2 0) production has been dominated by the Girdler-Sulphide (G-S) process, which suffers several intrinsic disadvantages that lead to high production costs. Processes based on hydrogen/water exchange have become more attractive with the development of proprietary wetproofed catalysts by AECL. One process that is synergistic with industrial hydrogen production by steam methane reforming (SMR), the combined industrial reforming and catalytic exchange (CIRCE) process, offers the best prospect for commercialization. SMRs are common globally in the oil upgrading and ammonia industries. To study the CIRCE process in detail, AECL, in collaboration with Air Liquide Canada, constructed a prototype CIRCE plant (PCP) in Hamilton, ON. The plant became fully operational in 2000 July and is expected to operate to at least the late fall of 2002. To date, plant operation has confirmed the adequacy of the design and the capability of enriching deuterium to produce heavy water without compromising hydrogen production. The proprietary wetproofed catalyst has performed as expected, both in activity and in robustness. (author)

  2. Prototype CIRCE plant - industrial demonstration of heavy water production from reformed hydrogen source

    International Nuclear Information System (INIS)

    Spagnolo, D.A.; Boniface, H.A.; Sadhankar, R.R.; Everatt, A.E.; Miller, A.I.; Blouin, J.

    2002-01-01

    Heavy water (D 2 0) production has been dominated by the Girdler-Sulphide (G-S) process, which suffers several intrinsic disadvantages that lead to high production costs. Processes based on hydrogen/water exchange have become more attractive with the development of proprietary wetproofed catalysts by AECL. One process that is synergistic with industrial hydrogen production by steam methane reforming (SMR), the Combined Industrial Reforming and Catalytic Exchange (CIRCE) process, offers the best prospect for commercialization. SMRs are common globally in the oil-upgrading and ammonia industries. To study the CIRCE process in detail, AECL, in collaboration with Air Liquide Canada, constructed a prototype CIRCE plant (PCP) in Hamilton, Ontario. The plant became fully operational in 2000 July and is expected to operate to at least late fall of 2002. To-date, plant operation has confirmed the adequacy of the design and the capability of enriching deuterium to produce heavy water without compromising hydrogen production. The proprietary wetproofed catalyst has performed as expected, both in activity and in robustness. (author)

  3. 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)

  4. 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

  5. Basic safety principles of KLT-40C reactor plants

    International Nuclear Information System (INIS)

    Beliaev, V.; Polunichev, V.

    2000-01-01

    The KLT-40 NSSS has been developed for a floating power block of a nuclear heat and power station on the basis of ice-breaker-type NSSS (Nuclear Steam Supply System) with application of shipbuilding technologies. Basic reactor plant components are pressurised water reactor, once-through coil-type steam generator, primary coolant pump, emergency protection rod drive mechanisms of compensate group-electromechanical type. Basic RP components are incorporated in a compact steam generating block which is arranged within metal-water shielding tank's caissons. Domestic regulatory documents on safety were used for the NSSS design. IAEA recommendations were also taken into account. Implementation of basic safety principles adopted presently for nuclear power allowed application of the KLT-40C plant for a floating power unit of a nuclear co-generation station. (author)

  6. Clinch river breeder reactor plant steam generator water quality

    International Nuclear Information System (INIS)

    Van Hoesen, D.; Lowe, P.A.

    1975-01-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

  7. Clinch river breeder reactor plant steam generator water quality

    Energy Technology Data Exchange (ETDEWEB)

    Van Hoesen, D; Lowe, P A

    1975-07-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

  8. The challenge of introducing high-temperature reactor plants onto the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

  9. The challenge of introducing high-temperature-reactor plants onto the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1988-01-01

    Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

  10. Reactor safety study applied to the Forsmark 3 Power Plant

    International Nuclear Information System (INIS)

    Ericsson, G.; Tiren, L.I.

    1978-01-01

    A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

  11. TMI-2 reactor vessel and balance of plant status

    International Nuclear Information System (INIS)

    Kuehn, G.A.

    1990-01-01

    In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

  12. A spectroscopic study of ion channels in a prototype inertial electrostatic confinement reactor

    International Nuclear Information System (INIS)

    Collis, S.; Khachan, J.

    2000-01-01

    Inertial Electrostatic Confinement (IEC) involves using a semi-transparent and negatively biased grid to accelerate light nuclei towards a common centre for the purpose of generating neutrons through fusion reactions. This project investigated the plasma properties in a small prototype IEC device that was operated using a relatively low grid bias in a discharge of hydrogen. Electrostatic lenses, which are the product of the geometry of the grid, create ion channels. Doppler shift spectroscopy was performed on the emission produced by charge exchange reactions in these channels. Using the spectra we obtained, we were able to determine energies, ratios of hydrogen species (H + :H 2 + :H 3 + ) and thermal properties of ions present in these channels. A discussion of results will be presented with particular emphasis on the implications of our findings to the construction of a portable neutron production device. (author)

  13. Reactor instrumentation and control in nuclear power plants in Germany

    International Nuclear Information System (INIS)

    Aleite, W.

    1993-01-01

    The pertinent legislation, guidelines and standards of importance for nuclear power plant construction as well as the relevant committees in Germany are covered. The impact of international developments on the German regulatory scene is mentioned. A series of 15 data sheets on reactor control, followed by 5 data sheets on instrumentation and control in nuclear power plants, which were drawn up for German plants, are compared and commented in some detail. Digitalization of instrumentation and control systems continues apace. To illustrate the results that can be achieved with a digitalized information system, a picture series that documents a plant test of behavior on simulated steam generator tube rupture is elaborately commented. An outlook on backfitting and upgrading applications concludes this paper. (orig.) [de

  14. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

  15. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  16. 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

  17. Balance of plant design issues for small reactors in Canada

    International Nuclear Information System (INIS)

    Harvel, G.; Meneley, D.

    2014-01-01

    Internationally, several companies are exploring design and development of Small Modular Reactors (SMR) ranging in power from 10 MWe to 300 MWe. While the designs are proceeding, the main issue at hand is finding a site for deployment of the first unit. Connection to existing well established grids is currently not competitive in part due to First of a Kind (FOAK) costs. As such, many vendors are exploring unique and remote applications where FOAK costs are not as significant a concern. One of the major assumptions in the design process usually followed is that the major effort needs to concentrate on reactor core development. While the reactor core is important, costs associated with the balance of plant and operations of the unit are likely to play an important role in the final decision of purchase. In this work, a series of conceptual designs is performed for the support systems of a small modular reactor by successive teams of undergraduate students working over semester long periods during a 3 year period. The goal of this process is to determine to what extent current technology for the balance of plant supports the development of a cost effective SMR. Each system is given to a team with an open set of criteria for design. At the completion of the design exercise, an open discussion with the teams is held regarding the staffing requirements for an SMR. The results are preliminary and reflect the open nature of the exercise. That said, the results indicate that for an SMR to be truly competitive, significant innovation is required in addressing the supporting systems of the plant. (author)

  18. Balance of plant design issues for small reactors in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Harvel, G.; Meneley, D., E-mail: Glenn.Harvel@uoit.ca, E-mail: dan.meneley@sympatico.ca [Univ. of Ontario Inst. of Tech.y, Oshawa, ON (Canada)

    2014-07-01

    Internationally, several companies are exploring design and development of Small Modular Reactors (SMR) ranging in power from 10 MWe to 300 MWe. While the designs are proceeding, the main issue at hand is finding a site for deployment of the first unit. Connection to existing well established grids is currently not competitive in part due to First of a Kind (FOAK) costs. As such, many vendors are exploring unique and remote applications where FOAK costs are not as significant a concern. One of the major assumptions in the design process usually followed is that the major effort needs to concentrate on reactor core development. While the reactor core is important, costs associated with the balance of plant and operations of the unit are likely to play an important role in the final decision of purchase. In this work, a series of conceptual designs is performed for the support systems of a small modular reactor by successive teams of undergraduate students working over semester long periods during a 3 year period. The goal of this process is to determine to what extent current technology for the balance of plant supports the development of a cost effective SMR. Each system is given to a team with an open set of criteria for design. At the completion of the design exercise, an open discussion with the teams is held regarding the staffing requirements for an SMR. The results are preliminary and reflect the open nature of the exercise. That said, the results indicate that for an SMR to be truly competitive, significant innovation is required in addressing the supporting systems of the plant. (author)

  19. DOE's annealing prototype demonstration projects

    International Nuclear Information System (INIS)

    Warren, J.; Nakos, J.; Rochau, G.

    1997-01-01

    One of the challenges U.S. utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels (RPVs). As a nuclear plant operates, its RPV is exposed to neutrons. For certain plants, this neutron exposure can cause embrittlement of some of the RPV welds which can shorten the useful life of the RPV. This RPV embrittlement issue has the potential to affect the continued operation of a number of operating U.S. pressurized water reactor (PWR) plants. However, RPV material properties affected by long-term irradiation are recoverable through a thermal annealing treatment of the RPV. Although a dozen Russian-designed RPVs and several U.S. military vessels have been successfully annealed, U.S. utilities have stated that a successful annealing demonstration of a U.S. RPV is a prerequisite for annealing a licensed U.S. nuclear power plant. In May 1995, the Department of Energy's Sandia National Laboratories awarded two cost-shared contracts to evaluate the feasibility of annealing U.S. licensed plants by conducting an anneal of an installed RPV using two different heating technologies. The contracts were awarded to the American Society of Mechanical Engineers (ASME) Center for Research and Technology Development (CRTD) and MPR Associates (MPR). The ASME team completed its annealing prototype demonstration in July 1996, using an indirect gas furnace at the uncompleted Public Service of Indiana's Marble Hill nuclear power plant. The MPR team's annealing prototype demonstration was scheduled to be completed in early 1997, using a direct heat electrical furnace at the uncompleted Consumers Power Company's nuclear power plant at Midland, Michigan. This paper describes the Department's annealing prototype demonstration goals and objectives; the tasks, deliverables, and results to date for each annealing prototype demonstration; and the remaining annealing technology challenges

  20. SEMPaC - an expert system prototype associated with safety parameter display system of a nuclear power plant

    International Nuclear Information System (INIS)

    Hirama, K.

    1989-01-01

    This work presents SEMPaC, an expert system prototype: it provides means to support diagnosis and to make decisions during abnormal transients that cause the trip of nuclear power plant. The system operation is associated with Safety Parameter Display System - SPDS that was recommended by U. S. Nuclear Regulatory Commission (NRC) after the Three-Mile Island (TMI) accident analysis. (author)

  1. Steam turbine chemistry in light water reactor plants

    International Nuclear Information System (INIS)

    Svoboda, Robert; Haertel, Klaus

    2008-01-01

    Steam turbines in boiling water reactor (BWR) and pressurized water reactor (PWR) power plants of various manufacturers have been affected by corrosion fatigue and stress corrosion cracking. Steam chemistry has not been a prime focus for related research because the water in nuclear steam generating systems is considered to be of high purity. Steam turbine chemistry however addresses more the problems encountered in fossil fired power plants on all volatile treatment, where corrosive environments can be formed in zones where wet steam is re-evaporated and dries out, or in the phase transition zone, where superheated steam starts to condense in the low-pressure (LP) turbine. In BWR plants the situation is aggravated by the fact that no alkalizing agents are used in the cycle, thus making any anionic impurity immediately acidic. This is illustrated by case studies of pitting corrosion of a 12 % Cr steel gland seal and of flow-oriented corrosion attack on LP turbine blades in the phase transition zone. In PWR plants, volatile alkalizing agents are used that provide some buffering of acidic impurities, but they also produce anionic decomposition products. (orig.)

  2. 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.)

  3. Repair and replacement of reactor internals for plant life extension

    International Nuclear Information System (INIS)

    Graae, T.

    1998-01-01

    Recent experience from early Swedish BWRs corroborate that all components in a nuclear power plant can be repaired or replaced with new ones. Oskarshamn 1 has gone through a thorough refurbishment project. A number of internals were repaired or replaced including the core shroud support which was welded to the bottom of the reactor pressure vessel. The project verifies that it is fully possible to carry out complicated inspection and repair work inside a nuclear pressure vessel which has been in operation for more than 20 years. Along with increased capacity factor, operating nuclear power plants get the financial conditions needed for extensive repair and modernization projects. Large power output leads to short pay-back times for the investments. The FENIX project at Oskarshamn 1 is such a project. There are utilities whose policy is to keep their plants in as-new condition for an unlimited length of time. (orig.)

  4. Nuclear power plant laundry drain treatment using membrane bio reactor

    International Nuclear Information System (INIS)

    Tsukamoto, Masaaki; Kohanawa, Osamu; Kinugasa, Atsushi; Ogawa, Naoki; Murogaki, Kenta

    2012-01-01

    In nuclear power plant, the radioactive effluent generated by washing the clothes worn in controlled area and the hand and shower water used at the controlled area are treated in laundry drain treatment system. Although various systems which treat such liquid waste preexist, the traditional treatment system has disadvantages such as high running cost and a large amount of secondary waste generation. To solve these matters, we have considered application of an activated sludge system, membrane bio reactor, which has been practically used in general industry. For nuclear power plant, the activated sludge system has been developed, tested in its adaptability and the adequacy has been proved. Some preexisting treatment systems have been replaced with this activated sludge system for the first time in a domestic nuclear power plant, and the renewal system is now in operation. The result is reported. (author)

  5. Studies on plant dynamics of sodium-cooled fast breeder reactors - verification of a plant model

    International Nuclear Information System (INIS)

    Schubert, B.

    1988-01-01

    For the analysis of sodium-cooled FBR safety and dynamics theoretical models are used, which have to be verified. In this report the verification of the plant model SSC-L is conducted by the comparison of calculated data with measurements of the experimental reactors KNK II and RAPSODIE. For this the plant model is extended and adapted. In general only small differences between calculated and measured data are recognized. The results are used to improve and complete the plant model. The extensions of the plant model applicability are used for the calculation of a loss of heat sink transient with reactor scram, considering pipes as passive heat sinks. (orig./HP) With 69 figs., 10 tabs [de

  6. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-06-15

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant.

  7. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    International Nuclear Information System (INIS)

    1986-06-01

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant

  8. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    Rust, J.H.; Clement, J.D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  9. Liquid radioactive waste processing system for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

  10. Annealing the reactor vessel at the Palisades Plant

    International Nuclear Information System (INIS)

    Fenech, R.A.

    1996-01-01

    In the way of background, Palisades was licensed in 1967 and went commercial in 1971. Jumping to two years ago, we faced at that time three issues that challenged our ability to operate to end-of-license, which would be 2007 without any extensions. The three items were regulatory performance, economic performance, and reactor vessel embrittlement. We had not been operating the plant with the kind of conservative decisions and with the kind of safety margins that one is expected to operate a plant in the United States at this time. Our economic performance was not satisfactory in that our capacity factor was low and our costs high. In the area of reactor vessel embrittlement, our analysis showed that we would reach the NRC screening criteria for embrittlement in the year 2004. Over the last two years, we have made significant improvements in the first two areas. Our decision-making has changed. Our performance, especially over the last year and a half, has been excellent. In addition, we have gotten our capacity factors up and our costs under control. Clearly, sustained performance is what is going to carry the day but from what we can see and from where we are, we are in more of a maintenance-of-performance than in a turn-around situation. On the other hand, in the area of reactor vessel embrittlement, about a year and a half ago we had a bit of a setback. We had taken material from retired steam generators that had welds identical to the welds in our reactor vessel. When we analyzed the welds from our steam generators, we were given some surprises about the chemistry makeup. When we applied the new information to our analysis, we changed the date on which we would reach our screening criteria from 2004 to late 1999

  11. Power Reactor Fuel Reprocessing Plant-2, Tarapur: a benchmark in Indian PHWR spent fuel reprocessing

    International Nuclear Information System (INIS)

    Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

    2017-01-01

    Power Reactor Fuel Reprocessing Plant-2 (PREFRE-2) is latest operating spent nuclear fuel reprocessing plant in India. This plant has improved design based on latest technology and feedback provided by the earlier plants. The design of PREFRE-2 plant is in five cycles of solvent extraction using TBP as extractant. The plant is commissioned in year 2011 after regulatory clearances

  12. Membrane bio-reactor for textile wastewater treatment plant upgrading.

    Science.gov (United States)

    Lubello, C; Gori, R

    2005-01-01

    Textile industries carry out several fiber treatments using variable quantities of water, from five to forty times the fiber weight, and consequently generate large volumes of wastewater to be disposed of. Membrane Bio-reactors (MBRs) combine membrane technology with biological reactors for the treatment of wastewater: micro or ultrafiltration membranes are used for solid-liquid separation replacing the secondary settling of the traditional activated sludge system. This paper deals with the possibility of realizing a new section of one existing WWTP (activated sludge + clariflocculation + ozonation) for the treatment of treating textile wastewater to be recycled, equipped with an MBR (76 l/s as design capacity) and running in parallel with the existing one. During a 4-month experimental period, a pilot-scale MBR proved to be very effective for wastewater reclamation. On average, removal efficiency of the pilot plant (93% for COD, and over 99% for total suspended solids) was higher than the WWTP ones. Color was removed as in the WWTP. Anionic surfactants removal of pilot plant was lower than that of the WWTP (90.5 and 93.2% respectively), while the BiAS removal was higher in the pilot plant (98.2 vs. 97.1). At the end cost analysis of the proposed upgrade is reported.

  13. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  14. Development of a Prototype Algal Reactor for Removing CO2 from Cabin Air

    Science.gov (United States)

    Patel, Vrajen; Monje, Oscar

    2013-01-01

    Controlling carbon dioxide in spacecraft cabin air may be accomplished using algal photobioreactors (PBRs). The purpose of this project was to evaluate the use of a commercial microcontroller, the Arduino Mega 2560, for measuring key photioreactor variables: dissolved oxygen, pH, temperature, light, and carbon dioxide. The Arduino platform is an opensource physical computing platform composed of a compact microcontroller board and a C++/C computer language (Arduino 1.0.5). The functionality of the Arduino platform can be expanded by the use of numerous add-ons or 'shields'. The Arduino Mega 2560 was equipped with the following shields: datalogger, BNC shield for reading pH sensor, a Mega Moto shield for controlling CO2 addition, as well as multiple sensors. The dissolved oxygen (DO) probe was calibrated using a nitrogen bubbling technique and the pH probe was calibrated via an Omega pH simulator. The PBR was constructed using a 2 L beaker, a 66 L box for addition of CO2, a micro porous membrane, a diaphragm pump, four 25 watt light bulbs, a MasterFiex speed controller, and a fan. The algae (wild type Synechocystis PCC6803) was grown in an aerated flask until the algae was dense enough to used in the main reactor. After the algae was grown, it was transferred to the 2 L beaker where CO2 consumption and O2 production was measured using the microcontroller sensor suite. The data was recorded via the datalogger and transferred to a computer for analysis.

  15. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

    Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

    2003-05-01

    In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

  16. Development of inelastic design method for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Takahashi, Yukio; Take, Kohji; Kaguchi, Hitoshi; Fukuda, Yoshio; Uno, Tetsuro.

    1991-01-01

    Effective utilization of inelastic analysis in structural design assessment is expected to play an important role for avoiding too conservative design of liquid metal reactor plants. Studies have been conducted by the authors to develop a guideline for application of detailed inelastic analysis in design assessment. Both fundamental material characteristics tests and structural failure tests were conducted. Fundamental investigations were made on inelastic analysis method and creep-fatigue life prediction method based on the results of material characteristics tests. It was demonstrated through structural failure tests that the design method constructed based on these fundamental investigations can predict failure lives in structures subjected to cyclic thermal loadings with sufficient accuracy. (author)

  17. Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

    Directory of Open Access Journals (Sweden)

    Chihyung Kim

    2016-04-01

    Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  18. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  19. Characterization of actinide physics specimens for the US/UK joint experiment in the Dounreay Prototype Fast Reactor

    International Nuclear Information System (INIS)

    Walker, R.L.; Botts, J.L.; Cooper, J.H.; Adair, H.L.; Bigelow, J.E.; Raman, S.

    1983-10-01

    The United States and the United Kingdom are engaged in a joint research program in which samples of the higher actinides are irradiated in the Dounreay Prototype Fast Reactor in Scotland. The purpose of the porogram is (1) to study the materials behavior of selected higher actinide fuels and (2) to determine the integral cross sections of a wide variety of the higher actinide isotopes. Samples of the actinides are incorporated in fuel pins inserted in the core. For the fuel study, the actinides selected are 241 Am and 244 Cm in the form of Am 2 O 3 , Cm 2 O 3 , and Am 6 Cm(RE) 7 O 21 , where (RE) represents a mixture of lanthanides. For the cross-section determinations, the samples are milligram quantities of actinide oxides of 248 Cm, 246 Cm, 244 Cm, 243 Cm, 243 Am, 241 Am, 244 Pu, 242 Pu, 241 Pu, 240 Pu, 239 Pu, 238 Pu, 237 Np, 238 U, 236 U, 235 U, 234 U, 233 U, 232 Th, 230 Th, and 231 Pa encapsulated in vanadium. Coincident with the irradiations, neutron flux and energy spectral measurements are made with vanadium-encapsulated dosimeter materials located within the same fuel pins

  20. The water desalination complex based on ABV-type reactor plant

    International Nuclear Information System (INIS)

    Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

    1997-01-01

    A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

  1. Irradiation of an uranium silicide prototype in RA-3 reactor; Irradiacion de un elemento combustible prototipo de siliciuro de uranio en el RA-3

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, R; Estrik, G; Notari, C [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    The factibility of irradiation of an uranium silicide (U{sub 3} Si{sub 2}) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U{sub 3} O{sub 8} as fissible material. The enrichment of both standard and prototype is the same: 20% U{sub 235} and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs.

  2. Plant life management processes and practices for heavy water reactors

    International Nuclear Information System (INIS)

    Kang, K.-S.; Cleveland, J.; Clark, C.R.

    2006-01-01

    In general, heavy water reactor (HWR) nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. Their decisions are depending on essentially business model. They involve the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. Continued plant operation, including operation beyond design life, called 'long term operation, depends, among other things, on the material condition of the plant. This is influenced significantly by the effectiveness of ageing management. Key attributes of an effective plant life management program include a focus on important systems, structure and components (SSCs) which are susceptible to ageing degradation, a balance of proactive and reactive ageing management programmes, and a team approach that ensures the co-ordination of and communication between all relevant nuclear power plant and external programmes. Most HWR NPP owners/operators use a mix of maintenance, surveillance and inspection (MSI) programs as the primary means of managing ageing. Often these programs are experienced-based and/or time-based and may not be optimised for detecting and/or managing ageing effects. From time-to-time, operational history has shown that this practice can be too reactive, as it leads to dealing with ageing effects (degradation of SSCs) after they have been detected. In many cases premature and/or undetected ageing cannot be traced back to one specific reason or an explicit error. The root cause is often a lack of communication, documentation and/or co-ordination between design, commissioning, operation or maintenance organizations. This lack of effective communication and interfacing frequently arises because, with the exception of major SSCs, such as the fuel channels or steam generators, there is a lack of explicit

  3. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  4. Extended fuel cycle operation for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Silvestri, G.J. Jr.

    1978-01-01

    A nuclear steam turbine power plant system having an arrangement therein for extended fuel cycle operation is described. The power plant includes a turbine connected at its inlet to a source of motive fluid having a predetermined pressure associated therewith. The turbine has also connected thereto an extraction conduit which extracts steam from a predetermined location therein for use in an associated apparatus. A bypass conduit is provided between a point upstream of the inlet and the extraction conduit. A flow control device is provided within the bypass conduit and opens when the pressure of the motive steam supply drops beneath the predetermined pressure as a result of reactivity loss within the nuclear reactor. Opening of the bypass conduit provides flow to the associated apparatus and at the same time provides an increased flow orifice to maintain fluid flow rate at a predetermined level

  5. Generation of spectrum compatible accelerograms for seismic analysis of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, T.; Chellapandi, P.; Chetal, S.C.

    2003-01-01

    For the seismic design of nuclear power plants, time history of earthquake ground motion is required basically to generate time histories at various floors of nuclear island as well as at the component support locations. From such time histories, floor response spectra (FRS) can be generated. The basic input is specified as site dependent response spectra (SDRS), from which a set of uncorrelated time histories is generated whose own response spectrum matches with the design response spectra. These time histories have got a great impact on the structural design and economy. For Kalpakkam, the site for PFBR, the seismic input is defined in terms of SDRS for various damping values and its shapes have been arrived already. Synthetic accelerograms have been generated such that the time-history generated response spectrum (THRS) closely matches the SDRS for 5% of critical damping. Time histories have been developed using CASTEM 2000, a multi purpose FE code. This paper deals with the generation methodology and their compliance with ASCE 4-98. (author)

  6. Leak detection in the primary reactor coolant piping of nuclear power plant by applying beam-microphone technology

    International Nuclear Information System (INIS)

    Kasai, Yoshimitsu; Shimanskiy, Sergey; Naoi, Yosuke; Kanazawa, Junichi

    2004-01-01

    A microphone leak detection method was applied to the inlet piping of the ATR-prototype reactor, Fugen. Statistical analysis results showed that the cross-correlation method provided the effective results for detection of a small leakage. However, such a technique has limited application due to significant distortion of the signals on the reactor site. As one of the alternative methods, the beam-microphone provides necessary spatial selectivity and its performance is less affected by signal distortion. A prototype of the beam-microphone was developed and then tested at the O-arai Engineering Center of the Japan Nuclear Cycle Development Institute (JNC). On-site testing of the beam-microphone was carried out in the inlet piping room of an RBMK reactor of the Leningrad Nuclear Power Plant (LNPP) in Russia. A leak sound imitator was used to simulate the leakage sound under the leakage flow condition of 1-3 gpm (0.23-0.7 m 3 /h). Analysis showed that signal distortion does not seriously affect the performance of this method, and that sound reflection may result in the appearance of ghost sound sources. The test results showed that the influences of sound reflection and background noise were smaller at the high frequencies where the leakage location could be estimated with an angular accuracy of 5deg which is the range of localization accuracy required for the leak detection system. (author)

  7. 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)

  8. Biodenitrification of gaseous diffusion plant aqueous wastes: stirred bed reactor

    International Nuclear Information System (INIS)

    Holland, M.E.

    1980-01-01

    Approximately 30 kilograms of nitrates per day are discarded in the raffinates (acid wastes) of the Portsmouth Gaseous Diffusion Plant's X-705 Uranium Recovery and Decontamination Facility. A biodenitrification process employing continuous-flow, stirred-bed reactors has been successfully used to remove nitrates from similar acid wastes at the Oak Ridge Y-12 Plant. Laboratory studies have been made at Portsmouth to characterize the X-705 raffinates and to test the stirred-bed biodenitrification process on such raffinates. Raffinates which had been previously characterized were pumped through continuous-flow, stirred-bed, laboratory-scale reactors. Tests were conducted over a period of 146 days and involved variations in composition, mixing requirements, and the fate of several metal ions in the raffinates. Tests results show that 20 weight percent nitrates were reduced to a target nitrate effluent concentration of 100 μg/ml with a 99.64 percent efficiency. However, the average denitrification rate achieved was only 33% of that demonstrated with the Y-12 stirred-bed system. These low rates were probably due to the toxic effects of heavy metal ions on the denitrifying bacteria. Also, most of the uranium in the raffinate feed remained in the biomass and calcite, which collected in the reactor. This could cause criticality problems. For these reasons, it was decided not to make use of the stirred-bed bioreactor at Portsmouth. Instead, the biodenitrification installation now planned will use fluidized bed columns whose performance will be the subject of a subsequent report

  9. Initiating Events for Multi-Reactor Plant Sites

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

  10. The ITER Fast Plant System Controller ATCA prototype Real-Time Software Architecture

    International Nuclear Information System (INIS)

    Carvalho, B.B.; Santos, B.; Carvalho, P.F.; Neto, A.; Boncagni, L.; Batista, A.J.N.; Correia, M.; Sousa, J.; Gonçalves, B.

    2013-01-01

    Highlights: ► High performance ATCA systems for fast control and data acquisition. ► IEEE1588 timing system and synchronization. ► Plasma control algorithms. ► Real-time control software frameworks. ► Targeted for nuclear fusion experiments with long duration discharges. -- Abstract: IPFN is developing a prototype Fast Plant System Controller (FPSC) based in ATCA embedded technologies dedicated to ITER CODAC data acquisition and control tasks in the sub-millisecond range. The main goal is to demonstrate the usability of the ATCA standard and its enhanced specifications for the high speed, very high density parallel data acquisition needs of the most demanding ITER tokamak plasma Instrumentation and Control (I and C) systems. This effort included the in-house development of a new family of high performance ATCA I/O and timing boards. The standard ITER software system CODAC Core System (CCS) v3.1, with the control based in the EPICS system does not cover yet the real-time requirements fulfilled by this hardware, so a new set of software components was developed for this specific platform, attempting to integrate and leverage the new features in CSS, for example the Multithreaded Application Real Time executor (MARTe) software framework, the new Data Archiving Network (DAN) solution, an ATCA IEEE-1588-2008 timing interface, and the Intelligent Platform Management Interface (IPMI) for system monitoring and remote management. This paper presents the overall software architecture for the ATCA FPSC, as well a discussion on the ITER constrains and design choices and finally a detailed description of the software components already developed

  11. The ITER Fast Plant System Controller ATCA prototype Real-Time Software Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, B.B., E-mail: bernardo@ipfn.ist.utl.pt [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Santos, B.; Carvalho, P.F.; Neto, A. [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Boncagni, L. [Associazione Euratom-ENEA sulla Fusione, Frascati Research Centre, Division of Fusion Physics, Frascati, Rome (Italy); Batista, A.J.N.; Correia, M.; Sousa, J.; Gonçalves, B. [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► High performance ATCA systems for fast control and data acquisition. ► IEEE1588 timing system and synchronization. ► Plasma control algorithms. ► Real-time control software frameworks. ► Targeted for nuclear fusion experiments with long duration discharges. -- Abstract: IPFN is developing a prototype Fast Plant System Controller (FPSC) based in ATCA embedded technologies dedicated to ITER CODAC data acquisition and control tasks in the sub-millisecond range. The main goal is to demonstrate the usability of the ATCA standard and its enhanced specifications for the high speed, very high density parallel data acquisition needs of the most demanding ITER tokamak plasma Instrumentation and Control (I and C) systems. This effort included the in-house development of a new family of high performance ATCA I/O and timing boards. The standard ITER software system CODAC Core System (CCS) v3.1, with the control based in the EPICS system does not cover yet the real-time requirements fulfilled by this hardware, so a new set of software components was developed for this specific platform, attempting to integrate and leverage the new features in CSS, for example the Multithreaded Application Real Time executor (MARTe) software framework, the new Data Archiving Network (DAN) solution, an ATCA IEEE-1588-2008 timing interface, and the Intelligent Platform Management Interface (IPMI) for system monitoring and remote management. This paper presents the overall software architecture for the ATCA FPSC, as well a discussion on the ITER constrains and design choices and finally a detailed description of the software components already developed.

  12. Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

    International Nuclear Information System (INIS)

    Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

    1999-09-01

    A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

  13. 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

  14. Prototype of generic server for wind power plants using IEC 61400-25 standard

    DEFF Research Database (Denmark)

    Olsen, Andreas K.; Ösdil, B.; Poulsen, Bjarne

    2006-01-01

    from interfaces for control and monitoring to a standardized and secure way of handling communication. An analysis focusing on isolating the necessary requirements has been carried out based on the IEC61400-25 in order to create a generic prototype which can be used by WPP vendors. The main...... communications interface of the prototype utilizes web services and the prototype developed is comprised of several independent modules to allow for the possibility of choosing a fully customizable setup by the end user. Configuration of the system needs to be done in a simple way, ensuring a flexible...... and reusable system, where different choices for the system can be added or left out depending on user specifications. From the requirements a prototype with the purpose of examining the key aspects of these definitions has been elaborated....

  15. 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

  16. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    2001-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  17. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    1999-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  18. Synthesis of the IRSN report on its analysis of the safety guidance package (DOrS) of the ASTRID reactor project. Safety guidance document for the ASTRID prototype: Referral to the GPR. Opinion related to the safety guidance document of the ASTRID reactor project. ASTRID prototype: Safety guidance document for the ASTRID prototype

    International Nuclear Information System (INIS)

    Lachaume, Jean-Luc; Niel, Jean-Christophe

    2013-01-01

    A first document indicates the improvement guidelines for the ASTRID project based on the French experience in the field of sodium-cooled fast neutron reactors, addresses the safety objectives as they are presented for the ASTRID project, discusses how the project includes a regulation and design referential, and how it addresses various aspects of the design approach (ranking and analysis of operation situations, defence in depth, use of probabilistic studies, safety classification and qualification to accidental situations, taking internal and external aggressions into account and taking severe accidents into account at the design level). It comments the guidelines related to the first two barriers, to main safety functions (control of reactivity and of reactor cooling, containment of radioactive and toxic materials), to dismantling, to R and D for safety support. A second document is a letter sent by the ASN to the GPR (permanent group of experts in charge of nuclear reactors) about the safety guidance document for the ASTRID prototype. The third document is the answer and contains comments and recommendations by this group about the content of this document, and therefore addresses the same topics as the first document. The last document defines the framework of the approach to this document

  19. Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Souza Mendes, J.E. de

    1987-01-01

    Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

  20. arXiv Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

    CERN Document Server

    Abreu, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B.C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L.N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

    2018-05-03

    The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration so...

  1. Dissipation of the reactor heat at the Savannah River Plant

    International Nuclear Information System (INIS)

    Neill, J.S.; Babcock, D.F.

    1971-10-01

    The effluent cooling water from the heat exchangers of the Savannah River nuclear reactors is cooled by natural processes as it flows through the stream beds, canals, ponds, and swamps on the plant site. The Langhaar equation, which gives the rate of heat removal from the water surface as a function of the surface temperature, air temperature, relative humidity, and wind speed, is applied satisfactorily to calculate the cooling that occurs at all temperature levels and for all modes of water flow. The application of this equation requires an accounting of effects such as solar heating, shading, mixing, staging, stratification, underflow, rainfall, the imposed heat load, and the rate of change in heat content of the body of water

  2. New designs of medium power WWER reactor plants

    International Nuclear Information System (INIS)

    Ryzhov, S.B.; Mokhov, V.A.; Nikitenko, M.P.; Chetverikov, A.E.; Veselov, D.O.; Shchekin, I.G.; Petrov, V.V.

    2010-01-01

    The task of constructing NPPs as the objects of regional power industry is included into the Federal Target Program on nuclear power technologies of new generation for the period till 2020. Such NPPs are considered as perspective sources of energy for solution of the problems concerning provision of electric energy, household and industrial heat to the regions with limited capabilities of the power grid. OKB 'GIDROPRESS' present the conceptual study of RP design for the Unit of 600 MW (el.) power, taking into account their long-term experience in the field of development and operation of WWER reactor plants. Practical implementation of WWER-600 and WWER-300 RP designs seems to be feasible: practice in manufacturing the main equipment is available; cooperation of design, scientific organizations and manufacturers of equipment; is established; basic design solutions for equipment are of reference character

  3. Seismic design criteria for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Morrone, A.; Bitner, J.L.; Sigal, G.B.

    1975-01-01

    The general criteria for seismic resistant design for structures, systems and components of the Clinch River Breeder Reactor Plant (CRBRP) are presented and discussed. Site dependency of the maximum ground accelerations for the Operating Basis Earthquake and the Safe Shutdown Earthquake is described from the viewpoint of historical records and geological and seismological studies for the CRBRP site. The respective ground response spectra are derived by normalization of the latest AEC Regulatory standard shapes to these maximum ground accelerations. Modeling and analytical techniques and requirements are given. In addition, loading conditions and categories, loading combinations, earthquake direction effects and allowable damping values are defined. A discussion of the testing criteria which considers both single and multiple frequency test motions, and basic test procedures for single frequency sine beat testing is presented. (U.S.)

  4. Model-Based Analysis and Efficient Operation of a Glucose Isomerization Reactor Plant

    DEFF Research Database (Denmark)

    Papadakis, Emmanouil; Madsen, Ulrich; Pedersen, Sven

    2015-01-01

    efficiency. The objective of this study is the application of the developed framework on an industrial case study of a glucose isomerization (GI) reactor plant that is part of a corn refinery, with the objective to improve the productivity of the process. Therefore, a multi-scale reactor model...... is developedfor use as a building block for the GI reactor plant simulation. An optimal operation strategy is proposed on the basis of the simulation results...

  5. A review of the U.K. fast reactor programme: March 1978

    International Nuclear Information System (INIS)

    Smith, R.D.

    1978-01-01

    The review of the UK fast reactor programme covers the description of Dounreay Fast Reactor shut down after seventeen years of successful operation; description of prototype fast reactor (PFR); core design parameters safety features and plant design for commercial demonstration fast reactor (CDFR). Engineering development is related to large sodium rigs, coolant circuit hydraulics and vibration, instrumentation and components. The subjects of interest are material development, sodium technology, fast reactor fuel, fuel cycle, reactor safety, reactor performance studies

  6. A review of the U.K. fast reactor programme: March 1978

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R D [United Kingdom Atomic Energy Authority, Risley (United Kingdom)

    1978-07-01

    The review of the UK fast reactor programme covers the description of Dounreay Fast Reactor shut down after seventeen years of successful operation; description of prototype fast reactor (PFR); core design parameters safety features and plant design for commercial demonstration fast reactor (CDFR). Engineering development is related to large sodium rigs, coolant circuit hydraulics and vibration, instrumentation and components. The subjects of interest are material development, sodium technology, fast reactor fuel, fuel cycle, reactor safety, reactor performance studies.

  7. Deployment of a Prototype Plant GFP Imager at the Arthur Clarke Mars Greenhouse of the Haughton Mars Project

    Directory of Open Access Journals (Sweden)

    Robert J. Ferl

    2008-04-01

    Full Text Available The use of engineered plants as biosensors has made elegant strides in the past decades, providing keen insights into the health of plants in general and particularly in the nature and cellular location of stress responses. However, most of the analytical procedures involve laboratory examination of the biosensor plants. With the advent of the green fluorescence protein (GFP as a biosensor molecule, it became at least theoretically possible for analyses of gene expression to occur telemetrically, with the gene expression information of the plant delivered to the investigator over large distances simply as properly processed fluorescence images. Spaceflight and other extraterrestrial environments provide unique challenges to plant life, challenges that often require changes at the gene expression level to accommodate adaptation and survival. Having previously deployed transgenic plant biosensors to evaluate responses to orbital spaceflight, we wished to develop the plants and especially the imaging devices required to conduct such experiments robotically, without operator intervention, within extraterrestrial environments. This requires the development of an autonomous and remotely operated plant GFP imaging system and concomitant development of the communications infrastructure to manage dataflow from the imaging device. Here we report the results of deploying a prototype GFP imaging system within the Arthur Clarke Mars Greenhouse (ACMG an autonomously operated greenhouse located within the Haughton Mars Project in the Canadian High Arctic. Results both demonstrate the applicability of the fundamental GFP biosensor technology and highlight the difficulties in collecting and managing telemetric data from challenging deployment environments.

  8. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    1997-01-01

    The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

  9. Reactor building design of nuclear power plant ATUCHA II, Argentina

    International Nuclear Information System (INIS)

    Rufino, R.E.; Hermann, E.R.; Richter, E.

    1984-01-01

    It is presented the civil engineering project carried out by the joint venture Hochtief - Techint-Bignoli (HTB) for the reactor building at the Atucha II power plant (PHWR of 745 MWe) in Buenos Aires. All the other civil projects at Atucha II are also being carried out by HTB. This building has the same general characteristics of the PWR plants developed by KWU in Germany, known for the spherical steel containment 56m in diameter. Nevertheless, it differs from those principally in the equipment lay-out and the remarkable foundation depth. From the basic engineering provided by ENACE, the joint venture has had to face the challenge of designing a tridimensional structure of large size. This has necessitated using simplified models which had to be superimposed, since the use of only one spatial mode would be highly inadequate, lacking the flexibility necessary to absorb the numerous modifications that this type of project undergoes during construction. In addition, this procedure has eliminated resorting to numerous and costly computer processings. (Author) [pt

  10. Studies on environment safety and application of advanced reactor for inland nuclear power plants

    International Nuclear Information System (INIS)

    Wei, L.; Jie, L.

    2014-01-01

    To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

  11. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    EL-Kafas, A.E.A.E.

    1996-01-01

    the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

  12. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  13. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    International Nuclear Information System (INIS)

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory

  14. Outline of nuclear power plant project of fast breeder reactor (FBR) 'Monju'

    International Nuclear Information System (INIS)

    Suzuki, Fumio; Higashide, Noboru

    1982-01-01

    The first step of the construction of the fast breeder prototype reactor ''Monju'' in Japan was taken by the declaration of agreement of Fukui Prefecture governor in May, 1982, and the succeeding approval by the cabinet council. In this paper, the outline of the project is described. The paper first introduces the development status of FBRs in the world, and next, describes the progress of planning the ''Monju'' project. Initially, ''Monju'' was invited by Shiraki region of Tsuruga City, but the move of village was not agreed. Thus, the center of the site was moved to ''Haseda'' about 1 km north-east. In addition, there were some limitations, and the layout of the facility and the design of public work structures including harbour construction were done under many limiting conditions. However, the site is rather in good conditions, and the design rationalization was attempted as far as possible within a range which does not affect the safety of the nuclear power plant. The outline of the examination related to public works and the siting conditions are reported, dividing them into topographic features, geology, meteorology, sea conditions, and natural environment. As the public works plan, site preparation, roads, harbour installations and water intake and discharge facility are also described. (Wakatsuki, Y.)

  15. Remote repairs and refurbishment of reactor internal structures of magnox plant

    International Nuclear Information System (INIS)

    Barnes, S.A.; Kelly, D.E.

    1992-01-01

    The original designers of the UK Magnox reactor plant made provision for the then perceived time dependent processes that could have influenced the operational life of the plant. Changes in graphite properties with irradiation, particularly dimensional change, were well understood and in-core samples were provided for subsequent laboratory examination to monitor the processes throughout plant life. The tendency towards embrittlement with irradiation of the steel of the reactor pressure vessels was also acknowledged and again in-core samples were provided for monitoring changes in materials properties in-service and thus provide data in support of structural analyses to sustain the reactor safety cases. (author)

  16. Outline of design, manufacturing and installation experience of pressure vessel structure for the prototype heavy water moderated boiling light water cooled reactor 'FUGEN'

    International Nuclear Information System (INIS)

    Shibato, Eizo; Oguchi, Isao; Kishi, Toshikazu; Kitagawa, Yuji

    1977-01-01

    After component installation completed in June 1977 and various functional tests to be conducted later, the prototype heavy water moderated, boiling light water cooled reactor ''FUGEN'' is scheduled to reach first criticality in March 1978. Since the pressure vessel of ''FUGEN'' is completely different from that of a light water reactor in structure and materials, through research and development work was carried out prior to fabrication and construction. Based on these studies, installation of the actual pressure vessel was completed. Functional tests are now under way. This article describes examples in which our research and development results are reflected on design, manufacture, and installation of the pressure vessel. Also it introduces noteworthy achievements relevant to production techniques in manufacture and installation. (auth.)

  17. Nuclear liability and research reactor fuel. A plant supplier's view

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Hetzmann, A.

    2000-01-01

    Contracts on Research Reactors are normally entered into by the owner and - very often - later user and the supplier of such plants. They are not concluded by the fuel supplier, except fuel supplier and plant supplier are identical. Thus, the fuel supplier mostly has no direct influence into the contract negotiations and the clauses which are eventually agreed upon between the parties. So has any other subcontractor for any other system or component. Any such subsupplier can and will negotiate a subsidiary supply contract (subcontract) with the supplier of the plant. The supplier drafts the related clauses so as to pass on to the subsupplier as many risks out of his contracts as possible. The subsupplier, on the other hand, tries to protect himself, tries to limit the risks he takes over, e.g. to the worth of his subcontract maximum. A critical issue in such negotiations is those concerning the risks the supplier had to accept and the subsupplier, although he may be responsible later for the risk changing to reality; i.e. the occurrence of a loss, cannot be hold liable for in full because the subcontract limits his liability, e.g. to the subcontract value or a certain delay penalty. A typical example for this conflict are delays of the entire project caused by one subsupplier. A very specific case in this context is the so-called nuclear liability. Nuclear liability means being hold responsible for the consequences or damages originating from a nuclear event in the plant. Those consequences or damages may be suffered by third parties, which are neither the owner/operator nor the supplier and result in a liability to such party (third party liability). Several of the aspects below may be related to the nuclear liability issue: The supplier often has its registered office not in the country where the plant is; The supplier may have far bigger assets than the owner/operator. The legal system of the supplier's country may be more favourable for enforcing claims of

  18. Integrated gasification gas combined cycle plant with membrane reactors: Technological and economical analysis

    International Nuclear Information System (INIS)

    Amelio, Mario; Morrone, Pietropaolo; Gallucci, Fausto; Basile, Angelo

    2007-01-01

    In the present work, the capture and storage of carbon dioxide from the fossil fuel power plant have been considered. The main objective was to analyze the thermodynamic performances and the technological aspects of two integrated gasification gas combined cycle plants (IGCC), as well as to give a forecast of the investment costs for the plants and the resulting energy consumptions. The first plant considered is an IGCC* plant (integrated gasification gas combined cycle plant with traditional shift reactors) characterized by the traditional water gas shift reactors and a CO 2 physical adsorption system followed by the power section. The second one is an IGCC M plant (integrated gasification gas combined cycle plant with membrane reactor) where the coal thermal input is the same as the first one, but the traditional shift reactors and the physical adsorption unit are replaced by catalytic palladium membrane reactors (CMR). In the present work, a mono-dimensional computational model of the membrane reactor was proposed to simulate and evaluate the capability of the IGCC M plant to capture carbon dioxide. The energetic performances, efficiency and net power of the IGCC* and IGCC M plants were, thus, compared, assuming as standard a traditional IGCC plant without carbon dioxide capture. The economical aspects of the three plants were compared through an economical analysis. Since the IGCC* and IGCC M plants have additional costs related to the capture and disposal of the carbon dioxide, a Carbon Tax (adopted in some countries like Sweden) proportional to the number of kilograms of carbon dioxide released in the environment was assumed. According to the economical analysis, the IGCC M plant proved to be more convenient than the IGCC* one

  19. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  20. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  1. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  2. High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area

  3. Transportable nuclear power plant T3C-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Ogloblin, B.G.; Gromov, B.F.

    1993-01-01

    Development and cultivating of districts in Siberia, North, Far East, Kamchatka and other remote or almost inaccessible district of the country depends to a large degree on their providing with power. The specific character of these districts imposes in turn a wide variety of special requirements upon the power sources. In particular, it is essential to provide the following; maximum manufacture availability of the whole equipment at the minimum volume of construction and installation work on operation site, high safety, longterm service life, ecologically, minimum scope of work on equipment in-service maintenance and inspection, etc. Taking into account the well-known difficulties connected with the delivery of conventional energy carriers to the above-mentioned districts and the situation with the alternative power sources, the application of the low-power nuclear plants (NPP) for these purposes looks definitely promising. Among the probable trends in creating the NPPs of this type as very promising is considered the possibility to apply the two-circuit reactor plant of the vessel type with the liquid lead as a primary coolant and free air as a secondary coolant and working medium in the open gas-turbine cycle. The nuclear plant T3C-M of improved safety with two of this type reactor plants with total electric power of 8 MW is developed by CDB of Machine Building with participation of several enterprises of St. Petersburg under the scientific leadership and is intended for generation of electric power and up to 4 Gcal/h of heat for populated areas and installations placed at long distance from the main electric power supply sources where it is difficult or non-efficient economically to deliver the conventional kinds of fuel. The main principles being laid as a basis when developing the proposed NPP will allow one to create mobile power sources which possess a high degree of safety and inherent self-protection

  4. Design study of plant system for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

    1986-11-01

    This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

  5. Risk-oriented analysis of the German prototype fast breeder reactor SNR-300: off-site accident consequence model and results of the study

    International Nuclear Information System (INIS)

    Bayer, A.; Ehrhardt, J.

    1984-01-01

    Accident off-site consequence calculations and risk assessments performed for the ''risk oriented analysis'' of the German prototype fast breeder reactor SNR-300 were performed with a modified version of the off-site accident consequence model UFOMOD. The modifications mainly relate to the deposition and resuspension processes, the ingestion model, and the dose factors. Consequence calculations at the site of Kalkar on the Rhine River were performed for 115 weather sequences in 36 wind directions. They were based on seven release categories evaluated for the SNR-300 with two different fueling strategies: plutonium from Magnox reactors only and plutonium from light water reactors and Magnox reactors. In parallel, the corresponding frequencies of occurrence are determined. The following results are generated: 1. complementary cumulative frequency distribution functions for collective fatalities and collective doses 2. expected values of the collective fatalities and collective doses as well as distance-dependent expected values of individual fatality 3. contributions of the different exposure pathways to fatalities with respect to the various organs. For comparison with the risk of a PWR-1300, calculations for the PWR-1300 of the ''German Risk Study'' were repeated with the same modified consequence model. Comparison shows that smaller risks result for the SNR-300. However, the confidence interval bandwidths obtained for the frequencies of the release categories for the SNR-300 are larger than those of the PWR-1300

  6. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  7. Chemical aspects of gadolinium nitrate as soluble nuclear poison in Savannah River Plant reactors

    International Nuclear Information System (INIS)

    Baumann, E.W.

    1978-01-01

    The aqueous solution chemistry of gadolinium nitrate was studied to identify conditions that interfere with successful cleanup of gadolinium in Savannah River Plant reactor systems. Injecting a gadolinium nitrate solution into the D 2 O coolant-moderator constitutes a supplementary mode of reactor shutdown. The resulting approximately 0.001M gadolinium nitrate solution is then deionized by recirculation through mixed-bed ion exchange resins before reactor operation is resumed

  8. Identification and management of plant aging and life extension issues for a liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    King, R.W.; Perry, W.H.

    1991-06-01

    Experimental Breeder Reactor 2 (EBR-2) is a pool-type sodium-cooled fast reactor that supports extensive experimental, test and demonstration programs while providing electrical power to the local grid. EBR-2 is a US Department of Energy Facility located at the Idaho National Engineering Laboratory and operated by Argonne National Laboratory (ANL). The current mission of EBR-2 is to serve as the operational prototype for the Integral Fast Reactor demonstration program. This mission and other programs require EBR-2 to operate reliability to a 40-year lifetime, a significant extension beyond the five to ten year life originally planned for the facility. The benefits of operating EBR-2 in the extended-life mode are important for providing long-term operational performance data for a sodium-cooled fast reactor that is not available elsewhere. Identification and preliminary assessment of potential life-limiting factors indicate that, with appropriate consideration given in the design phase, the sodium-cooled plant has potential for a very long operational lifetime. Achievement of a 40-year lifetime with high reliability is important not only for achieving the near-term goals of the EBR-2/IFR programs, but for the advancement of the liquid-metal-cooled reactor concept to the demonstration/commercialization phase. Key features make extended-life operation feasible based on the use of sodium as the primary coolant: low-pressure, high thermal capacity primary system and a low-pressure secondary system requiring no active valves; and limited corrosion of components. 2 refs

  9. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

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

    International Nuclear Information System (INIS)

    Pan Xiren; Zhang Chen.

    1988-01-01

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

  11. Catalogue and classification of technical safety rules for light-water reactors and reprocessing plants

    International Nuclear Information System (INIS)

    Bloser, M.; Fichtner, N.; Neider, R.

    1975-08-01

    This report on the cataloguing and classification of technical rules for land-based light-water reactors and reprocessing plants contains a list of classified rules. The reasons for the classification system used are given and discussed

  12. 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.

  13. 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

  14. General aspects of CAREM-25 reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

    2000-01-01

    CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

  15. General Aspects of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

    2000-01-01

    CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

  16. Nuclear energy center site survey reactor plant considerations

    International Nuclear Information System (INIS)

    1976-05-01

    The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers

  17. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  18. Rock siting of nuclear power plants from a reactor safety standpoint. Status report October 1974

    International Nuclear Information System (INIS)

    1975-01-01

    The aim of this study is to clearify the advantages and disadvantages of an underground nuclear power plant from a reactor safety point of view, compared to a plant above ground. Principles for the technical design of a rock sited BWR nuclear power plant is presented. Also questions of sabotage and closing down the plant at the end of the operational period are treated. (K.K.)

  19. 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

  20. Alstom's Chemical Looping Combustion Prototype for CO2 Capture from Existing Pulverized Coal-Fired Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Andrus, Jr., Herbert E. [Alstom Power Inc., Windsor, CT (United States); Chiu, John H. [Alstom Power Inc., Windsor, CT (United States); Edberg, Carl D. [Alstom Power Inc., Windsor, CT (United States); Thibeault, Paul R. [Alstom Power Inc., Windsor, CT (United States); Turek, David G. [Alstom Power Inc., Windsor, CT (United States)

    2012-09-30

    Alstom’s Limestone Chemical Looping (LCL™) process has the potential to capture CO2 from new and existing coal-fired power plants while maintaining high plant power generation efficiency. This new power plant concept is based on a hybrid combustion- gasification process utilizing high temperature chemical and thermal looping technology. This process could also be potentially configured as a hybrid combustion-gasification process producing a syngas or hydrogen for various applications while also producing a separate stream of CO2 for use or sequestration. The targets set for this technology is to capture over 90% of the total carbon in the coal at cost of electricity which is less than 20% greater than Conventional PC or CFB units. Previous work with bench scale test and a 65 kWt Process Development Unit Development (PDU) has validated the chemistry required for the chemical looping process and provided for the investigation of the solids transport mechanisms and design requirements. The objective of this project is to continue development of the combustion option of chemical looping (LCL-C™) by designing, building and testing a 3 MWt prototype facility. The prototype includes all of the equipment that is required to operate the chemical looping plant in a fully integrated manner with all major systems in service. Data from the design, construction, and testing will be used to characterize environmental performance, identify and address technical risks, reassess commercial plant economics, and develop design information for a demonstration plant planned to follow the proposed Prototype. A cold flow model of the prototype will be used to predict operating conditions for the prototype and help in operator training. Operation of the prototype will provide operator experience with this new technology and performance data of the LCL-C™ process, which will be applied to the commercial design and economics and plan for a future demonstration

  1. Power unit with GT-MHR reactor plant for electricity production and district heating

    International Nuclear Information System (INIS)

    Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

    2000-01-01

    Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

  2. Plant dynamics analyses of fast reactor concept: RAPID-A without any control rod

    International Nuclear Information System (INIS)

    Kambe, Mitsuru

    1996-01-01

    Plant dynamics analyses of a fast reactor concept RAPID-A without any control rod have been demonstrated in case of reactor startup and sudden change of the primary flow rate. RAIP-A concept involves Lithium Expansion Module (LEM) for inherent reactivity feedback, Lithium Injection Module (LIM) for inherent ultimate shutdown and Lithium Release Module (LRM) for automated reactor startup. LEM consists of Quick-LEM and Slow-LEM. Slow-LEM provides with moderate reactivity addition as decreasing temperature. Quick-LEM assures quick negative reactivity feedback as increasing temperature. Plant dynamics analyses revealed that reactor power is nearly proportional to the primary flow rate even if the flow rate increases suddenly. Fully automated reactor startup from the subcritical condition has been attempted by inserting reactivity at a constant rate by LRM. Allowable rate of reactivity addition has been obtained in respect to Quick-LEM reactivity worth. (author)

  3. Instrumentation and control for reactor power setback in PFBR

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Vasal, Tanmay; Nagaraj, C.P.; Madhusoodanan, K.

    2013-01-01

    In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

  4. Ensuring the operational safety of nuclear power plants with WWER reactors

    International Nuclear Information System (INIS)

    Shasharin, G.A.; Veretennikov, G.A.; Abagyan, A.A.; Lesnoj, S.A.

    1984-01-01

    At the start of 1983, 27 nuclear power producing units with reactor facilities of the WWER type were in operation in the Soviet Union and other countries. In 1982 the average load factor for nuclear power plants with WWER reactors was 73 per cent. There was not a single nuclear accident or even damage with any significant radiation consequences in the WWER reactors during the entire period of their operation. The most modern nuclear power plants with WWER-440 and WWER-1000 reactors meet all present-day international requirements. Safe operation of the plants is achieved by a variety of measures, the most important of which include: procedures for increasing the reliability of plant equipment and systems; ensuring exact compliance with plant operating instructions; ensuring reliable operation of plant safety systems; action directed towards maintaining the skills of plant personnel at a level adequate to ensure the taking of proper action during transient processes and accident situations. The paper discusses concrete steps for ensuring safe nuclear power plant operation along these lines. In particular, measures such as the following are described: the use of a system for collecting and processing information on equipment failures and defects; the development and introduction of methods of early defect diagnosis; the performance of complex testing of safety systems; the training of highly skilled personnel for nuclear power plants at educational combines and at teaching and training centres making use of simulators; arranging accident-prevention training and special instruction for personnel. (author)

  5. Design of a Prototype of Water Purification by Plasma Technology as the Foundation for an Industrial Wastewater Plant

    International Nuclear Information System (INIS)

    Barillas, L

    2015-01-01

    In order to mitigate the contamination of water sources due to the spill of sewage without any kind of treatment, mainly generated by the industrial sector; a prototype of water purification by plasma technology has been designed. The prototype will transform liquid water into plasma to eliminate the pathogens from the water, due to their exposure to ultraviolet radiation, electric fields and shock waves, which aid in the destruction of pollutants. The sewage will be accelerated at high speed to convert it into a liquid-gas mixture in order to transform it into plasma, which is achieved when the electrical discharge (of the type dielectric barrier discharge or DBD) is applied to the water by means of high voltage electrodes, from a source of alternating current (AC). Subsequently, the mixture slows down to be return into liquid phase and obtain clean water, all of these without a significantly rise of temperature. The device also has an automatic power control system. Finally, a short feasibility study was conducted in order to use this type of water cleaner in the future as a basis for a treatment plant of industrial waste water, so it comes to replace the current secondary and tertiary treatments used among the industry. It is intended that this new system will be more efficient and cheaper than the current waste water treatments. (paper)

  6. Design of a Prototype of Water Purification by Plasma Technology as the Foundation for an Industrial Wastewater Plant

    Science.gov (United States)

    Barillas, L.

    2015-03-01

    In order to mitigate the contamination of water sources due to the spill of sewage without any kind of treatment, mainly generated by the industrial sector; a prototype of water purification by plasma technology has been designed. The prototype will transform liquid water into plasma to eliminate the pathogens from the water, due to their exposure to ultraviolet radiation, electric fields and shock waves, which aid in the destruction of pollutants. The sewage will be accelerated at high speed to convert it into a liquid-gas mixture in order to transform it into plasma, which is achieved when the electrical discharge (of the type dielectric barrier discharge or DBD) is applied to the water by means of high voltage electrodes, from a source of alternating current (AC). Subsequently, the mixture slows down to be return into liquid phase and obtain clean water, all of these without a significantly rise of temperature. The device also has an automatic power control system. Finally, a short feasibility study was conducted in order to use this type of water cleaner in the future as a basis for a treatment plant of industrial waste water, so it comes to replace the current secondary and tertiary treatments used among the industry. It is intended that this new system will be more efficient and cheaper than the current waste water treatments.

  7. Operating results obtained in a nuclear power plant with a sensor surveillance prototype

    International Nuclear Information System (INIS)

    Jacquot, J.P.; Poujol, A.; Beaubatie, J.; Ciaramitaro, W.

    1983-03-01

    Surveillance methods have been validated and specific equipment have been built to measure the response time of sensors from a nuclear power plant protection channel. The reason of the choice of this parameter is twofold: the sensor response time is representative of the sensor physical status and is also part of the overall channel response time. Two surveillance methods are used: noise analysis (by AR or PSD modeling), and loop current step response (for resistance thermometer detectors only). The methods were validated on test facilities and on nuclear power plants. Two test equipments were built and tested on plants. Results are represented and conclusions are drawn on the feasibility of such methods for sensor surveillance [fr

  8. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Executive summary: main report

    International Nuclear Information System (INIS)

    1975-10-01

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

  9. Pilot plant experience on high-level waste solidification and design of the engineering prototype VERA

    Energy Technology Data Exchange (ETDEWEB)

    Guber, W; Diefenbacher, W; Hild, W; Krause, H; Schneider, E; Schubert, G

    1972-11-01

    In the present paper the solidification process for highly active waste solutions as developed in the Karlsruhe Nuclear Research Center is presented. Its principal steps are: denitration, calcination in a spray calciner operated with superheated steam, melting of the calcine with appropriate additives to borosilicate glass in an induction-heated melting furnace. The operational experiences gained so far in the inactive 1:1 pilot plant are reported. Furthermore, a description is given of the projected multi-purpose experimental facility VERA 2 which is provided for processing the highly active waste solutions from the first German reprocessing plant WAK.

  10. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

  11. Application of fault tree methodology to modeling of the AP1000 plant digital reactor protection system

    International Nuclear Information System (INIS)

    Teolis, D.S.; Zarewczynski, S.A.; Detar, H.L.

    2012-01-01

    The reactor trip system (RTS) and engineered safety features actuation system (ESFAS) in nuclear power plants utilizes instrumentation and control (IC) to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations. During normal operating conditions, various plant parameters are continuously monitored to assure that the plant is operating in a safe state. In response to deviations of these parameters from pre-determined set points, the protection system will initiate actions required to maintain the reactor in a safe state. These actions may include shutting down the reactor by opening the reactor trip breakers and actuation of safety equipment based on the situation. The RTS and ESFAS are represented in probabilistic risk assessments (PRAs) to reflect the impact of their contribution to core damage frequency (CDF). The reactor protection systems (RPS) in existing nuclear power plants are generally analog based and there is general consensus within the PRA community on fault tree modeling of these systems. In new plants, such as AP1000 plant, the RPS is based on digital technology. Digital systems are more complex combinations of hardware components and software. This combination of complex hardware and software can result in the presence of faults and failure modes unique to a digital RPS. The United States Nuclear Regulatory Commission (NRC) is currently performing research on the development of probabilistic models for digital systems for inclusion in PRAs; however, no consensus methodology exists at this time. Westinghouse is currently updating the AP1000 plant PRA to support initial operation of plants currently under construction in the United States. The digital RPS is modeled using fault tree methodology similar to that used for analog based systems. This paper presents high level descriptions of a typical analog based RPS and of the AP1000 plant digital RPS. Application of current fault

  12. Experience gathered from the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type

    International Nuclear Information System (INIS)

    Pastorini, A.; Belinco, C.G.; El Bis, E.D.; Sacchi, M.A.; Mayans, C.O.; Martin Ghiselli, A.; Marcora, G.R.

    1990-01-01

    This work describes the needs to materialize the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type, under special conditions, from the Fabrication Pilot Plant sited at the Constituyentes Atomic Center and the Ezeiza Atomic Center, for its subsequent analysis at the High Pressure Experimental Loop. The special conditions under which the transport has been made responded to the fact that the prototype presents a fragile adjustment between rods and separators, necessary to be preserved. (Author) [es

  13. Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

    International Nuclear Information System (INIS)

    Toft, P.; Borring, J.; Adolph, E.

    1988-01-01

    A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

  14. Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

    International Nuclear Information System (INIS)

    Anderson, J.W.

    1980-01-01

    The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

  15. Plasma properties in a large-volume, cylindrical and asymmetric radio-frequency capacitively coupled industrial-prototype reactor

    International Nuclear Information System (INIS)

    Lazović, Saša; Puač, Nevena; Spasić, Kosta; Malović, Gordana; Petrović, Zoran Lj; Cvelbar, Uroš; Mozetič, Miran; Radetić, Maja

    2013-01-01

    We have developed a large-volume low-pressure cylindrical plasma reactor with a size that matches industrial reactors for treatment of textiles. It was shown that it efficiently produces plasmas with only a small increase in power as compared with a similar reactor with 50 times smaller volume. Plasma generated at 13.56 MHz was stable from transition to streamers and capable of long-term continuous operation. An industrial-scale asymmetric cylindrical reactor of simple design and construction enabled good control over a wide range of active plasma species and ion concentrations. Detailed characterization of the discharge was performed using derivative, Langmuir and catalytic probes which enabled determination of the optimal sets of plasma parameters necessary for successful industry implementation and process control. Since neutral atomic oxygen plays a major role in many of the material processing applications, its spatial profile was measured using nickel catalytic probe over a wide range of plasma parameters. The spatial profiles show diffusion profiles with particle production close to the powered electrode and significant wall losses due to surface recombination. Oxygen atom densities range from 10 19 m −3 near the powered electrode to 10 17 m −3 near the wall. The concentrations of ions at the same time are changing from 10 16 to the 10 15 m −3 at the grounded chamber wall. (paper)

  16. Inventory of nuclear power plants and research reactors temporary or definitively stopped in industrialized countries

    International Nuclear Information System (INIS)

    Clauzon, J.; Vaubert, B.

    1984-12-01

    This paper presents data and information on the end of the life of nuclear reactors. One deals more particularly with installations of industrialized countries. This report gives the motivations which have involved the definitive shut down of nuclear power plants and of research reactors in the concerned countries. A schedule of definitive reactor shutdowns is presented. Then, one deals with nuclear power plants of which the construction has been stopped. The reasons of these situations are also given. The temporary difficulties met during the construction or the starting of nuclear power plants these last years are mentioned. Most times, there are economical or political considerations, or safety reasons. Finally, the nuclear power plants stopped for more than two years are mentioned [fr

  17. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  18. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens, E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas

    2015-07-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

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

    International Nuclear Information System (INIS)

    Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

    2001-01-01

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

  20. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    International Nuclear Information System (INIS)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens

    2015-01-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  1. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  2. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  3. Architectural prototyping

    DEFF Research Database (Denmark)

    Bardram, Jakob Eyvind; Christensen, Henrik Bærbak; Hansen, Klaus Marius

    2004-01-01

    A major part of software architecture design is learning how specific architectural designs balance the concerns of stakeholders. We explore the notion of "architectural prototypes", correspondingly architectural prototyping, as a means of using executable prototypes to investigate stakeholders...

  4. 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

  5. International breeder reactor development

    International Nuclear Information System (INIS)

    Traube, K.

    1976-01-01

    For more than a decade, sodium cooled breeder reactors have now been in the focus of advanced nuclear power development in the major industrialized countries. In the sixties, a total of seven small experimental nuclear power stations were commissioned. Two of these have been shut down in the meantime, the others continue to work satisfactorily, their main purpose being the development of fuel elements. The years 1972-1974 saw the commissioning of the prototype power stations in the 300 MWe power category in France, the United Kingdom and the Soviet Union. Presently, other experimental reactors are under construction in the Federal Republic of Germany, Italy, Japan, the United States, plus another Soviet 600 MWe prototype reactor and the SNR 300 DeBeNeLux prototype at Kalkar. A comparison of the technological features either implemented or planned in the prototype and experimental power plants and of their fuel elements reveals a remarkable similarity in the basic concepts pursued in different countries. The two types of breeder reactors, viz. the loop and the pool types, show a closer resemblance to each other than do pressurized and boilling water reactors. The growing awareness of administrative problems emerging in the approaching phase of the introduction of large breeder power stations in a number of European countries has recently led to a streamlining effort in the structure of industries and to tentative steps towards international cooperation on a broad basis. (orig.) [de

  6. The safety characteristics of the HTR 500 reactor plant

    International Nuclear Information System (INIS)

    Wachholz, W.

    1987-01-01

    The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

  7. 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

  8. Feasibility of a single-purpose reactor plant for district heating in Finland

    International Nuclear Information System (INIS)

    Tarjanne, R.; Vuori, S.; Eerikaeinen, L.; Saukkoriipi, L.

    A feasibility study of a single-purpose reactor for district heating is presented. The reactor chosen is of an ordinary pressurized water reactor type with a thermal output of 100 to 200 MW. Primary circuit steam generators employed in ordinary PWR's are replaced by water-water heat exchangers. For safety reasons an intermediate circuit separates the primary from the network water. The conditions of the district heating systems in Finland were taken into account, which led to the choice of an average temperature of 160 0 C for the reactor coolant and a pressure of 13.5 bar. This, coupled with minimal control requirements helped design a considerably simple reactor plant. On condition, the reactor satisfies the basic heat demand in a district heating system, the effective annual full-power operation time of the heating reactor is from 5000 h to 7000 h. Economic comparisons indicated that the heating reactor may be competitive if the operation period is of this order. As the reactor has to be sited near the heat consumption area for reasons of economy, the safety aspects are of major importance and may in themselves preclude the realization of the heating idea. (author)

  9. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    International Nuclear Information System (INIS)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP

  10. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP.

  11. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  12. Lessons learned from the licensing process for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Dickson, P.W.; Clare, G.H.

    1991-01-01

    This paper presents the experience of licensing a specific liquid-metal fast breeder reactor (LMFBR), the Clinch River Breader Reactor Plant (CRBRP). It was a success story in that the licensing process was accomplished in a very short time span. The actions of the applicant and the actions of the US Nuclear Regulatory Commission (NRC) in response are presented and discussed to provide guidance to future efforts to license unconventional reactors. The history is told from the perspective of the authors. As such, some of the reasons given for success or lack of success are subjective interpretations. Nevertheless, the authors' positions provided them an excellent viewpoint to make these judgements. During the second phase of the licensing process, they were the CRBRP Technical Director and the Licensing Manager, respectively, for the Westinghouse Electric Corporation, the prime contractor for the reactor plant

  13. Development of operator thinking model and its application to nuclear reactor plant operation system

    International Nuclear Information System (INIS)

    Miki, Tetsushi; Endou, Akira; Himeno, Yoshiaki

    1992-01-01

    At first, this paper presents the developing method of an operator thinking model and the outline of the developed model. In next, it describes the nuclear reactor plant operation system which has been developed based on this model. Finally, it has been confirmed that the method described in this paper is very effective in order to construct expert systems which replace the reactor operator's role with AI (artificial intelligence) systems. (author)

  14. PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1979-10-01

    The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined

  15. PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1979-10-01

    The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

  16. 14th Biennial conference on reactor operating experience plant operations: The human element

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

  17. Chemical reactor for a PUREX reprocessing plant of 200Kg U/day capacity

    International Nuclear Information System (INIS)

    Oliveria Lopes, M.J. de.

    1974-03-01

    Dissolution of spent reactor fuels in Purex process is studied. Design of a chemical reactor for PWR elements, 3% enriched uranium dioxide with zircaloy cladding, for a 200Kg/day uranium plant is the main objective. Chop-leach process is employed and 7.5M nitric acid is used. Non-criticality was obtained by safe geometry and checked by spectrum homogeneous calculus and diffusion codes. Fuel cycle is considered and decladding and dissolution are treated more accurately

  18. Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

    Science.gov (United States)

    Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

    2017-03-01

    The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

  19. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  20. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  1. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

    2000-01-01

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form (uranium oxide), which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design

  2. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

    2000-04-01

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  3. Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2000-06-07

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of {approx}11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide (U{sub 3}O{sub 8})], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

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

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

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

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

    International Nuclear Information System (INIS)

    1975-10-01

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

  6. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

  7. Systematic simulation of a tubular recycle reactor on the basis of pilot plant experiments

    Energy Technology Data Exchange (ETDEWEB)

    Paar, H; Narodoslawsky, M; Moser, A [Technische Univ., Graz (Austria). Inst. fuer Biotechnologie, Mikrobiologie und Abfalltechnologie

    1990-10-10

    Systematic simulatiom may decisively help in development and optimization of bioprocesses. By applying simulation techniques, optimal use can be made of experimental data, decreasing development costs and increasing the accuracy in predicting the behavior of an industrial scale plant. The procedure of the dialogue between simulation and experimental efforts will be exemplified in a case study. Alcoholic fermentation of glucose by zymomonas mobilis bacteria in a gasified turbular recycle reactor was studied first by systematic simulation, using a computer model based solely on literature data. On the base of the results of this simulation, a 0.013 m{sup 3} pilot plant reactor was constructed. The pilot plant experiments, too, were based on the results of the systematic simulation. Simulated and experimental data were well in agreement. The pilot plant experiments reiterated the trends and limits of the process as shown by the simulation results. Data from the pilot plant runs were then used to improve the simulation model. This improved model was subsequently used to simulate the performances of an industrial scale plant. The results of this simulation are presented. They show that the alcohol fermentation in a tubular recycle reactor is potentially advantageous to other reactor configurations, especially to continuous stirred tanks. (orig.).

  8. Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Waloch, M.L.

    1977-01-01

    In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

  9. Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

    International Nuclear Information System (INIS)

    Brown, N.R.; Revankar, S.T.; Seker, V.; Downar, Th.J.

    2010-01-01

    In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO 3 decomposition section of the SI cycle is proposed and modelled. Since SO 3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear

  10. Nuclear power plant monitoring method by neural network and its application to actual nuclear reactor

    International Nuclear Information System (INIS)

    Nabeshima, Kunihiko; Suzuki, Katsuo; Shinohara, Yoshikuni; Tuerkcan, E.

    1995-11-01

    In this paper, the anomaly detection method for nuclear power plant monitoring and its program are described by using a neural network approach, which is based on the deviation between measured signals and output signals of neural network model. The neural network used in this study has three layered auto-associative network with 12 input/output, and backpropagation algorithm is adopted for learning. Furthermore, to obtain better dynamical model of the reactor plant, a new learning technique was developed in which the learning process of the present neural network is divided into initial and adaptive learning modes. The test results at the actual nuclear reactor shows that the neural network plant monitoring system is successfull in detecting in real-time the symptom of small anomaly over a wide power range including reactor start-up, shut-down and stationary operation. (author)

  11. Uranium-236 in light water reactor spent fuel recycled to an enriching plant

    International Nuclear Information System (INIS)

    de la Garza, A.

    1977-01-01

    The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

  12. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    Science.gov (United States)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  13. EPRI's nuclear power plant instrumentation and control program and its applicability to advanced reactors

    International Nuclear Information System (INIS)

    Naser, J.; Torok, R.; Wilkinson, D.

    1997-01-01

    I ampersand C systems in nuclear power plants need to be upgraded over the lifetime of the plant in a reliable and cost-effective manner to replace obsolete equipment, to reduce O ampersand M costs, to improve plant performance, and to maintain safety. This applies to operating plants now and will apply to advanced reactors in the future. The major drivers for the replacement of the safety, control, and information systems in nuclear power plants are the obsolescence of the existing hardware and the need for more cost-effective power production. Competition between power producers is dictating more cost-effective power production. The increasing O ampersand M costs to maintain systems experiencing obsolescence problems is counter to the needs for more cost-effective power production and improved competitiveness. This need for increased productivity applies to government facilities as well as commercial plants. Increasing competition will continue to be a major factor in the operation of both operating plants and advanced reactors. It will continue to dictate the need for improved productivity and cost-effectiveness. EPRI and its member nuclear utilities are working together on an industry wide I ampersand C Program to address I ampersand C issues and to develop cost-effective solutions. A majority of the I ampersand C products and demonstrations being developed under this program will benefit advanced reactors in both the design and operational phases of their life cycle as well as it will benefit existing plants. 20 refs

  14. Study on dual plant concept for the next generation boiling water reactors

    International Nuclear Information System (INIS)

    Sato, Takashi; Oikawa, Hirohide

    1999-01-01

    The paper presents the study results on the basic concept of dual BWRs. For the convenience, we call the concept here as Trial Study on BWR dual concept (TSBWR dual). The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The TSBWR dual is a plant concept of dual BWRs contained in a same secondary containment building. The plant output is from 2 x l,350 MWe up to 2 x 1,700 MWe. This concept is mainly aiming at safety improvement and cost savings of the next generation BWRs. The TSBWR dual has two RPVs and two dry wells (DW). It has, however, only one wet well (WW) and only one R/B. The WW and the R/B are shared by the dual reactors. The operating floor is also shared by the two reactors. The TSBWR dual has both passive safety systems and active safety systems. They are also shared between the two reactors. A lot of sharing between the dual reactors enables significant cost savings accompanied by the power increase up to 3,400 MWe. Although the TSBWR dual consists of two reactors, the simplified cylindrical configuration of the key structures and reduction of the R/B height can minimize the plant construction period. The TSBWR dual provides a concept with which we can challenge to construct a dual BWR plant in the near future. (author)

  15. 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

  16. Solar Radiation Distribution inside a Greenhouse Prototypal with Photovoltaic Mobile Plant and Effects on Flower Growth

    Directory of Open Access Journals (Sweden)

    Andrea Colantoni

    2018-03-01

    Full Text Available The diffusion of renewable energy requires the search for new technologies useful for obtaining good energy and production efficiency. Even if the latter is not always easy to obtain, the integration of photovoltaic panels on the roof of greenhouses intended for floriculture can represent an alternative. The present paper evaluates climatic conditions inside a greenhouse, in which 20% of its roof surface has been replaced with mobile photovoltaic (PV panels. The PV system implemented in this study can vary the light energy collection surface in relation to the degree of insolation. The aim is to observe the shading effects of the PV system on the growth of several varieties of flowers (iberis, mini-cyclamens and petunias to ensure the use of solar energy as an income integration deriving from floricultural production. In fact, in agronomic terms, it has ensured: (i to be able to shade the underlying environment in most lighting conditions; and (ii to let through more light when it is required for the needs of crop plants or in cloudy weather. Results have described the distribution of solar radiation, variability of temperature and humidity and lighting in a solar year and the observed outcomes on floristic production.

  17. An evaluation of the fluid-elastic instability for Intermediate Heat Exchanger of Prototype Sodium-cooled fast Reactor

    International Nuclear Information System (INIS)

    Cho, Jaehun; Kim, Sungkyun; Koo, Gyeonghoi

    2014-01-01

    The sodium-cooled fast reactor (SFR) module consists of the vessel, containment vessel, head, rotating plug (RP), upper internal structure (UIS), intermediate heat exchanger (IHX), decay heat exchanger (DHX), primary pump, internal structure, internal components and reactor core. The IHXs transfer heat from the radioactive sodium coolant (primary sodium) in the primary heat transport system to the nonradioactive sodium coolant (secondary sodium) in the intermediate heat transport system. Each sodium flows like Fig. 1. Primary sodium flows inside of tube and secondary sodium flows outside. During transferring heat two sodium to sodium, the fluid-elastic instability is occurred among tube bundle by cross flow. Large amplitude vibration occurred by the fluid-elastic instability is caused such as crack and wear of tube. Thus it is important to decrease the fluid-elastic instability in terms of a safety. The purpose of this paper is to evaluate the fluid-elastic instability for tube bundle in the IHX following ASME code. This paper evaluated the fluid-elastic instability of tube bundle in the SFR IHX. According evaluation results, the fluid-elastic instability of IHX tube bundle is occurred. A installing an additional TSP under the upper tubesheet can decrease a probability of fluid-elastic instability. If a location of an additional TSP does not exceed tube length to become a 750 mm, tube bundle of IHX is safety from the fluid-elastic instability

  18. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  19. Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

    International Nuclear Information System (INIS)

    Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

    1987-01-01

    The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

  20. 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

  1. The working lifetime of nuclear power plants and new types of power reactors

    International Nuclear Information System (INIS)

    Bataille, Ch.; Birraux, C.

    2003-01-01

    The report on the working lifetime of nuclear power plants and new reactor types, by Mr Christian Bataille, deputy for the Nord, and Mr Claude Birraux, deputy for Haute-Savoie as well as President of the Office, supplements the studies carried out by the Parliamentary Office on the Safety of Nuclear Installations and Radioactive Wastes: it examines the remaining working life of the EDF nuclear power plants and the current status of projects that might, if circumstances were right, replace the reactors at present in service. The report investigates the different physical and other factors that influence the ageing of nuclear power plants and tackles the question of whether the design life of 40 years could be exceeded in practice. The whole issue of French nuclear power plant is put in perspective and compared with the situation of nuclear plants in Finland, Sweden, Germany and the United States, from the technical and regulatory standpoints. Believing that any attempt to optimise the working lifetime of the power plants currently in service must be accompanied by simultaneous moves aimed at their replacement, Messrs. Christian Bataille and Claude Birraux go on to review in detail the various light water reactor projects being proposed around the world for completion by 2015, as developments of existing models, in particular the EPR reactor of Framatome ANP, characterised by its competitiveness. They suggest that a first such reactor should be built as quickly as possible. Describing the other nuclear systems being investigated by research organisations not only in France but also in the United States and Sweden, Mrs. Christian Bataille and Claude Birraux review the objectives of these and the circumstances in which they might be developed, which would be unlikely to be before 2035 in view of the technological problems to be overcome and the industrial demonstration plants that would be needed

  2. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

  3. Draft report on compilation of generic safety issues for light water reactor nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    A generally accepted approach to characterizing the safety concerns in nuclear power plants is to express them as safety issues which need to be resolved. When such safety issues are applicable to a generation of plants of a particular design or to a family of plants of similar design, they are termed generic safety issues. Examples of generic safety issues are those related to reactor vessel embrittlement, control rod insertion reliability or strainer clogging. The safety issues compiled in this document are based on broad international experience. This compilation is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. Refs.

  4. Draft report on compilation of generic safety issues for light water reactor nuclear power plants

    International Nuclear Information System (INIS)

    1997-07-01

    A generally accepted approach to characterizing the safety concerns in nuclear power plants is to express them as safety issues which need to be resolved. When such safety issues are applicable to a generation of plants of a particular design or to a family of plants of similar design, they are termed generic safety issues. Examples of generic safety issues are those related to reactor vessel embrittlement, control rod insertion reliability or strainer clogging. The safety issues compiled in this document are based on broad international experience. This compilation is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. Refs

  5. 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

  6. Nuclear power plants without the risk of a reactor disaster

    International Nuclear Information System (INIS)

    Merz, L.

    1980-01-01

    Nowadays everybody believes that all nuclear power plants are an inevitable risk for the life of the people in its environment; this is a prejudice. This article points out that it is possible to plan nuclear power plants in a way that no deaths in the population have to be feared, even in the heaviest possible disturbances. In order to realise such nuclear power plants, it is necessary either to improve the cooling of the fuel rods by passive additional safety measures of which one unlimitedly efficient thus excluding a nuclear melting accident or to improve the safety containers so that it remains sufficiently tight. (orig./HP) [de

  7. A prototype expert system to support the development of a fault-tree analysis software for nuclear reactor safety

    International Nuclear Information System (INIS)

    Mesko, L.

    1990-01-01

    The project called EMERIS is designed to provide a material testing nuclear reactor and experimental loops with a software for the 'acquisition, evaluation and archivation of measured data during the operation of the experimental facility'. The project which gives job a team has a duration of two years and involves three Vax compatible TPA-type computers and many smaller computers for data digitalization and graphical workstations. The detailed description of the project is not the task of the paper. One of its modules, however, plays an important role in the considerations. Namely the module for distrubance analysis (DA) which is planned to perform a rule based on-line evaluation of numerous predefined fault trees in an expert system like environment

  8. Analysis of power ramp rate and minimum power controllability of the MMS model for a plant dynamics analysis of a Prototype SFR

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Kim, Dehee; Joo, Hyungkook; Lee, Taeho

    2014-01-01

    A full plant dynamic model was developed for a prototype SFR using the Modular Modeling System (MMS). It includes the modeling of various subsystems such as the neutronics, primary and intermediate sodium systems of the NSSS, steam and water systems of the BOP, BOP controls, and the supervisory plant controls. The NSSS model is subdivided into component models, such as a Core, IHXs, Pumps, SGs, and the rest of the NSSS loop model. The BOP model is subdivided into a steam subsystem, feedwater subsystem, and preheater subsystem. Plant transient tests were performed to study the operational considerations. It includes varying the power ramp rate and studying the controllability at minimum power. Plant transient tests were performed to study operational considerations by using the MMS model for a prototype SFR. It includes varying the power ramp rate, studying the controllability at the minimum power set point. At a power ramp rate of higher than 2%, the steam temperature has a large deviation from the target. As the power set point decreases, the PHTS hot leg temperature and steam temperature tend to have higher deviations. After further refinement of the MMS model, it can be useful for developing the plant operation logics of the prototype SFR

  9. Applicability of base-isolation R ampersand D in non-reactor facilities to a nuclear reactor plant

    International Nuclear Information System (INIS)

    Seidensticker, R.W.; Chang, Y.W.

    1990-01-01

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R ampersand D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R ampersand D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R ampersand D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R ampersand D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs

  10. Pressurized water reactor nuclear power plant. Environmental characterization information report

    International Nuclear Information System (INIS)

    1981-01-01

    The typical plant chosen for characterization is a 10000-MWe nameplate rating with wet-natural-draft cooling towers and modern radwaste control and processing equipment. The process, plant operating parameters, resources needed, and the environmental residuals and products associated with the power plant are presented. Annual resource usage and pollutant discharges are shown in English and metric units, assuming an annual plant capacity factor of 70%. In addition to annual quantities, the summary table gives quantities in terms of 10 12 Btu (about 293 million kWh) of electrical energy produced for comparison among energy processes. Supporting information and calculation procedures for the data are given. Thirteen environmental points of interest are discussed individually. Cost information, typical radioactive releases, and use of cooling ponds as an alternative cooling method are discussed in appendixes. A glossary and list of acronyms and abbreviations are provided

  11. General view about reactor safety nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Gasparian, A.E.; Silva, D.E.; Salvatore, J.E.L.; Lima, J.M. de

    1991-01-01

    In this paper the authors describe the principles and goals that have guided, as well as the methods that have been used by the National Commission of Nuclear Energy (CNEN) to set forth measures aiming at providing safety to the Brazilian nuclear power plants. The status of the licensing process of these power plants is shown. The performance and the results obtained so far in relation to the nuclear safety are also described. (author)

  12. Prediction of hydrogen distribution in the reactor building in CANDU6 plant

    International Nuclear Information System (INIS)

    Jin, Y.; Song, Y.

    2008-01-01

    The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

  13. Commissioning and Operational Experience in Power Reactor Fuel Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Pradhan, S., E-mail: spradhan@barctara.gov.in [Tarapur Based Reprocessing Plant, Bhabha Atomic Research Centre, Tarapur (India)

    2014-10-15

    After completing design, construction, commissioning, operation and maintenance experience of the reprocessing plants at Tarapur, Mumbai and Kalpakkam a new reprocessing plant is commissioned and put into operation at BARC, Tarapur since 2011. Subsequent to construction clearance, commissioning of the plant is taken in many steps with simultaneous review by design and safety committees. In spite of vast experience, all the staff was retrained in various aspects of process and utility operations and in operation of innovative changes incorporated in the design. Operating personnel are licensed through an elaborate procedure consisting of various check lists followed by personnel interview. Commissioning systems were divided in sub-systems. Sub-systems were commissioned independently and later integrated testing was carried out. For commissioning, extreme operating conditions were identified in consultation with designers and detailed commissioning procedures were made accordingly. Commissioning was done in different conditions to ensure safety, smooth operation and maintainability. Few modifications were carried out based on commissioning experience. Technical specifications for operation of the plant are made in consultation with designers and reviewed by safety committees. Operation of the plant was carried out after successful commissioning trials with Deep Depleted Uranium (DDU). Emergency operating procedures for each design basis accident were made. Performance of various systems, subsystems are quite satisfactory and the plant has given very good capacity factor. (author)

  14. Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Carl, H.; Kubis, K.

    1979-01-01

    Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

  15. The reactor accident in the Three Mile Island-2 reactor plant. (Harrisburg, USA)

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The presentation of the accident development is based on the bulletin of the Atomic Industrial Forum (AIF) from the 6th of April, 1979. In addition, there are some short pieces of information, No. 14 and 15 of the association for reactor security as well as written and verbal information of the firms Brown, Boveri and Co/Babcock-Brown Boveri Reaktor GmbH (BBC/BBR). (orig.) [de

  16. Selection of power plant elements for future reactor space electric power systems

    International Nuclear Information System (INIS)

    Buden, D.; Bennett, G.A.; Copper, K.

    1979-09-01

    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

  17. Nuclear reactor plant with a small gas-cooled HT reactor accommodated in a steel pressure vessel

    International Nuclear Information System (INIS)

    Schoening, J.; Elter, C.

    1986-01-01

    The plant has a small HT reactor and an He/He heat exchanger situated above this, with preferably two parallel circulating blowers connected after it. It also has at least one post-shutdown heat removal system, which is situated after the He/He heat exchanger in the direction of flow and which always has the total quantity of primary helium flowing through it. In one version of the design, the heat exchanger consists of two concentric bundles of helices connected after one another, which have primary helium flowing in one direction and secondary helium in the opposite direction. (orig./HP) [de

  18. Nuclear reactor plant with a gas-cooled nuclear reactor situated in a cylindrical prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Becker, G.; Elter, C.; Fritz, R.; Rautenberg, J.; Schoening, J.; Stracke, W.

    1986-01-01

    A simplified construction of the nuclear reactor plant with a guarantee of great safety is achieved by the auxiliary heat exhangers, which remove the post-shutdown heat in fault situations, being arranged in the wellknown way in pairs above one another in a vertical shaft. The associated auxiliary blowers are situated at the top for the upper auxiliary heat exchangers and at the bottom for the lower auxiliary heat exchangers. The cold gas is taken from the lower auxiliary blowers through a parallel gas pipe laid in concrete, which enters the vertical shaft concerned in the area of the cold gas pipe. (orig./HP) [de

  19. Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

    International Nuclear Information System (INIS)

    Lear, G.; Erickson, P.B.

    1975-01-01

    Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

  20. Outline of construction planning on No. 2 Reactor of the Shika Nuclear Power Plant

    International Nuclear Information System (INIS)

    Nakagawa, Tetsuro; Kadoki, Shuichi; Kubo, Tetsuji

    1999-01-01

    The Hokuriku Electric Co., Ltd. carries out the expansion of the Shika Nuclear Power Plant No.2 (ABWR) to start its in March 2006. It is situated in north neighboring side of No. 1 reactor under operation at present, and its main buildings are planned to position a reactor building at mountain side and a turbine building at sea side as well as those in the No. 1 reactor. And, cooling water for steam condenser was taken in from an intake opening built at north side of the lifting space situated at the front of the power plant, and discharged into seawater from a flashing opening positioned about 600 m offing. Here were described on outline of main civil engineering such as base excavation engineering, concrete caisson production, oceanic establishment engineering, and facility for steam condenser, and characteristics of the engineering. (G.K.)

  1. Limiting factor analysis of high availability nuclear plants (boiling water reactors). Final report

    International Nuclear Information System (INIS)

    Frederick, L.G.; Brady, R.M.; Shor, S.W.W.; McCusker, J.T.; Alden, W.M.; Kovacs, S.

    1979-08-01

    The pertinent results are presented of a 16-month study conducted for Electric Power Research Institute by General Electric Company, Bechtel Power Corporation, and Philadelphia Electric Company. The study centered around the Peach Bottom 2 Atomic Power Station, but also included limited study of operations at 20 additional operating boiling water reactors. The purpose of the study was to identify and evaluate key factors limiting plant availability, and to identify potential improvements for eliminating or alleviating those limitations. The key limiting factors were found to be refueling activities; activities related to the reactor fuel; reactor scrams; activities related to 20 operating systems or major components; delays due to radiation, turbid water during refueling operations, facilities/working conditions, and dirt/foreign material; and general maintenance/repair of valves and piping. Existing programs to reduce the effect on plant unavailability are identified, and suggestions for further action are made

  2. Reactor control device for controlling load of nuclear power plant

    International Nuclear Information System (INIS)

    Hirota, Tadakuni; Yokoyama, Terukuni; Masuda, Jiro.

    1981-01-01

    Purpose: To improve the load follow-up capacity of a nuclear reactor by automatically controlling the width of the not-sensing band of a control rod inserting and removing discriminator circuit. Constitution: When load control operations such as automatic load control, automatic frequency control, governor free operation and so forth are conducted, the width of a not sensing band of a control rod inserting and removing discriminator circuit is ao automatically controlled that the not sensing band width may return to ordinary value in a normal operation by avoiding the fast repetition of inserting and removing control rods by increasing the width of the insensing band if the period of a control deviation signal produced due to the variation in the load is quickly repeated and varied in correspondence to the control deviation signal. That is, a circuit for varying the insensing band of the control circuit for driving a control mechanism is provided to reduce the amount of driving the control rods in a load control operation and to reduce the strain of the power distribution of the nuclear reactor, thereby improving the load control capacity. (Yoshihara, H.)

  3. Validation of BN Reactor Plant Long-Term Operation

    International Nuclear Information System (INIS)

    Vilensky, O.; Vasilyev, B.; Kaidalov, V.

    2013-01-01

    The BN RP operation life time is mainly determined by resource of non-replaceable equipment. The new standard (RD) “Procedure of strength analysis for main components of sodium cooled fast neutron reactor plants” was developed to validate structure strength in view of radiation effects and degradation of material properties within the time period up to 300000 hours and under irradiation, as well as development of postulated crack-like defects. Using this RD, the extension of operation life of BN-600 reactor non-replaceable components from 30 to 45 years, as well as strength and durability of the most loaded non-replaceable components of BN-800 RP under construction were validated for the specified 45-year operation life. Wider application of steel 16Cr-11Ni-3Mo refers to new decisions in BN-1200 RP design that allow increasing of operation life of the most loaded non-replaceable components up to 60 years. High-chromium steel 12Cr-Ni-Mo-V-Nb is a new material, which was proposed for SG design to increase the operation life up to 30 years. In addition, the austenitic steels 18Cr-9Ni and 16Cr-11Ni-3Mo are now under upgrading for future application of them in commercial BN-1200 RP. To provide additional long-term reliable and safe operation of BN-1200 RP equipment and pipelines, it is planned to develop and implement the lifetime operational monitoring system

  4. Reactor power control systems in nuclear power plants

    International Nuclear Information System (INIS)

    Nakajima, Kazuo.

    1980-01-01

    Purpose: To enable power control by automatic control rod operation based on the calculated amounts of operation for the control rods determined depending on a power set value from reactor operators or on power variation amounts from other devices. Constitution: When an operator designates an automatic selection by way of a control rod operation panel, automatic signals are applied to a manual-automatic switching circuit and the mode judging circuit of a rod pattern control device. Then, mode signals such as for single operation, load setting, load following and the like produced by the operator are judged in a circuit, wherein a control rod pattern operation circuit calculates the designation for the control rods and the operation amounts for the control rods depending on the designated modes and automatic control is conducted for the control rods by a rod position control circuit, a rod drive control device and the like connected at a rod position monitor device. The reactor power is thus controlled automatically to reduce the operator's labours. The automatic power control can also be conducted in the same manner by the amount of power variations applied to the device from the external device. (Yoshino, Y.)

  5. 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

  6. Dynamic simulation of a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  7. Technology programs in support of advanced light water reactor plants: Construction

    International Nuclear Information System (INIS)

    Eichen, E.P.

    1989-10-01

    Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule duration, and to simplify design. This document discusses construction approaches. 77 refs., 5 figs., 6 tabs

  8. Technology programs in support of advanced light water reactor plants: Construction

    International Nuclear Information System (INIS)

    Eichen, E.P.

    1987-12-01

    Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses successes and problems in construction. 49 refs., 16 figs., 8 tabs

  9. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

  10. Technology programs in support of advanced light water reactor plants: Construction

    International Nuclear Information System (INIS)

    Eichen, E.P.

    1989-01-01

    Under Contract No. AC03-86SF16565, Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses design requirements. 36 refs., 57 figs., 56 tabs

  11. Seawater desalination plant using nuclear heating reactor coupled with MED process

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10~200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

  12. Temperature conditions of foundation plates under nuclear power plant reactor compartments

    International Nuclear Information System (INIS)

    Ehsaulov, S.L.

    1990-01-01

    Method for calculation of temperature conditions for foundation plates under reactor compartments located in the main building, used in construction of the second stage of the Kostroma nuclear power plant, is considered. The obtained calculation data can be used for determining the most suitable period of concrete placement, composition, initial temperature, manufacturing technology and ways of delivery of concrete mixture

  13. Study on reactor power change and ambiguous control of third Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Wang Gongzhan

    2006-01-01

    The phenomenon of the average power reduction during long term full power operating in Third Qinshan nuclear power plant is analyzed . According to the basic conclusions of reactor power fluctuating derived by probability statistic and calculation the corresponding ambiguous control project is proposed. The operating performance could be achieved by the present controlling project is predicted additionally. (authors)

  14. Seawater desalination plant using nuclear heating reactor coupled with MED process

    International Nuclear Information System (INIS)

    Wu Shaorong; Dong Duo; Zhang Dafang; Wang Xiuzhen

    2000-01-01

    A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. this seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. the intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m 3 /d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented

  15. Dynamic simulation of a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant

  16. Pump selection and application in a pressurized water reactor electric generating plant

    International Nuclear Information System (INIS)

    Kitch, D.M.

    1985-01-01

    Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

  17. Techniques for processing remote field eddy current signals from bend regions of steam generator tubes of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thirunavukkarasu, S. [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Rao, B.P.C., E-mail: bpcrao@igcar.gov.in [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Jayakumar, T.; Raj, Baldev [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2011-04-15

    Steam generator (SG) is one of the most critical components of sodium cooled fast breeder reactor. Remote field eddy current (RFEC) technique has been chosen for in-service inspection (ISI) of these ferromagnetic SG tubes made of modified 9Cr-1Mo steel (Grade 91). Expansion bends are provided in the SGs to accommodate differential thermal expansion. During ISI using RFEC technique, in expansion bend regions, exciter-receiver coil misalignment, bending stresses, probe wobble and magnetic permeability variations produce disturbing noise hindering detection of defects. Fourier filtering, cross-correlation and wavelet transform techniques have been studied for noise reduction as well as enhancement of RFEC signals of defects in bend regions, having machined grooves and localized defects. Performance of these three techniques has been compared using signal-to-noise ratio (SNR). Fourier filtering technique has shown better performance for noise reduction while cross-correlation technique has resulted in significant enhancement of signals. Wavelet transform technique has shown the combined capability of noise reduction and signal enhancement and resulted in unambiguous detection of 10% of wall loss grooves and localized defects in the bend regions with SNR better than 7 dB.

  18. South Africa's nuclear model: A small and innovative reactor is seen as the model for new electricity plants. The project is nearing the starting blocks

    International Nuclear Information System (INIS)

    Ferreira, Tom

    2004-01-01

    Although nuclear power generation has by far the best safety and environmental record of any technology in general use, it has for many years been unable to make any meaningful inroads into the wall of negative perceptions that have arisen against it. But sentiments are changing rapidly on a global scale. The flare-up of oil prices is a sobering reminder of the volatility in the energy market, the exhaustibility of fossil fuels and the urgent need for stable, reliable, non-polluting sources of electrical power that are indispensable to a modern industrial economy. Today, new types of nuclear plants are prized, and South Africa is moving ahead. The State energy provider, Eskom, is internationally regarded as the leader in the field of the Pebble Bed Modular Reactor (PBMR) technology, a 'new generation' nuclear power plant. A decision on the PBMR project's future is on the near horizon. Should approvals be received in the coming months to proceed to the project's next phase, construction of the PBMR demonstration plant will start in 2006, in which case the reactor will start in 2010 and handed over to the client, Eskom, in 2011. Eskom has conditionally undertaken to purchase the first commercial units. Pebble bed reactors are small, about one-sixth the size of most current nuclear plants. Multiple PBMRs can share a common control center and occupy an area of no more than three football fields. More specifically, the PBMR is a helium-cooled, graphite moderated high temperature reactor (HTR). The concept is based on experience in the UK, United States and particularly Germany where prototype reactors were operated successfully between the late 1960s and 1980s. Although it is not the only high-temperature, gas-cooled nuclear reactor being developed in the world, the South African project is internationally regarded as a front-runner. The South African PBMR includes unique and patented technological innovations which make it particularly competitive. The Chief Executive

  19. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  20. Collaborative Prototyping

    DEFF Research Database (Denmark)

    Bogers, Marcel; Horst, Willem

    2014-01-01

    of the prototyping process, the actual prototype was used as a tool for communication or development, thus serving as a platform for the cross-fertilization of knowledge. In this way, collaborative prototyping leads to a better balance between functionality and usability; it translates usability problems into design......This paper presents an inductive study that shows how collaborative prototyping across functional, hierarchical, and organizational boundaries can improve the overall prototyping process. Our combined action research and case study approach provides new insights into how collaborative prototyping...... can provide a platform for prototype-driven problem solving in early new product development (NPD). Our findings have important implications for how to facilitate multistakeholder collaboration in prototyping and problem solving, and more generally for how to organize collaborative and open innovation...

  1. Plant availability design aspects of Korean next generation reactor

    International Nuclear Information System (INIS)

    Woo Sang Lim; Ha Chung Beak

    1998-01-01

    The purpose of this paper is to describe the KNGR design concepts adopted for reducing forced outages and refueling outages, and current design changes, to assess their availability impacts compared to existing domestic nuclear power plants, and then to identify design directions for next design stage. (author)

  2. Increase of efficiency of plant materials heat treatment in tubular reactors

    Directory of Open Access Journals (Sweden)

    A. V. Golubkovich

    2016-01-01

    Full Text Available In agriculture products of pyrolysis of plant materials in the form of waste of the main production can be applied as a source of heat and electric power. Besides, their use prevents ecological pollution of the soil and the atmosphere. Pyrolysis plants can be used for work with tubular reactors anywhere. Due to them farmers can dry grain, using waste heat of diesel generators, heatgenerators, boiler plants and receiving thus gaseous products, liquid and firm fractions. A technology based on cyclic and continuous plant mass movement by a piston in a pipe from a loading site to a place of unloading of a firm phase consistently through cameras of drying, pyrolysis, condensation of gaseous products. Exhaust furnace gases with a temperature up to 600 degrees Celsius are given countercurrent material movement from a power equipment. The gaseous, liquid and firm products from the pyrolysis camera are used for heat and electric power generation. Calculation of parameters of subdrying and pyrolysis cameras is necessary for effective and steady operation of the tubular reactor. The authors determined the speed of raw materials movement, and also duration of drying and pyrolysis in working chambers. An analysis of a simplified mathematical model of process was confirmed with results of experiments. Models of heat treatment of wet plant materials in tubular reactors are worked out on a basis of equality of speeds of material movement in the reactor and distribution of a temperature front in material on radius. The authors defined estimated characteristic for determination of tubular reactor productivity and size of heat, required for drying and pyrolysis.

  3. An organic profile of a pressurised water reactor secondary plant

    International Nuclear Information System (INIS)

    Eeden, Nestor van; Stwayi, Mandisibuntu; Gericke, Gerhard

    2012-01-01

    Make-up water addition to the steam/water cycle at Koeberg Nuclear Power Station usually results in a corresponding increase of the chloride concentration in the steam generator blowdown system. During plant transients, when higher than normal make-up is required to the secondary plant, the concentration of chloride occasionally exceeds the limiting value for the station chemistry performance indicator. Irrespective of this, the demineralised water make-up supply tanks, which are routinely analysed for chloride, are within all recognised acceptable standards for secondary water make-up and therefore these tanks do not initially appear to be the source of chloride contamination. Water treatment at the plant relies essentially on ion exchange, which has been proven to be very effective in removing inorganic ionic species such as chloride. Organic compounds are less effectively removed by ion exchange and may pass through the treatment system, and these organics can reside undetected in the make-up water tanks. Historically, the elevated chloride concentration following high system make-up has been attributed to chlorinated organic compounds known as trihalomethanes being present in the make-up water tanks, but no rigorous study had been undertaken. As it has been assumed that the majority of chloride in the secondary system originates from the make-up water organic impurities, it was considered important to confirm this by compiling an organic profile of the secondary plant. The use of organic additives was also taken into account in the profile. This work has confirmed the contribution from trihalomethanes and has also found that other organochlorides contribute even more significantly to the overall chloride inventory of the secondary plant. (orig.)

  4. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    International Nuclear Information System (INIS)

    Woo, H.H.; Lu, S.C.

    1981-01-01

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

  5. Halden reactor project

    International Nuclear Information System (INIS)

    1980-01-01

    The research programme at the Halden Project is focused on the following three areas: 1. In-core behavior of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2. Prediction, surveillance and control of fuel and core performance for which models of fuel and core behavior are developed. 3. Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

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

    International Nuclear Information System (INIS)

    1975-11-01

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

  7. Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

    International Nuclear Information System (INIS)

    Chang H. Oh; Eung Soo Kim; Steven Sherman

    2008-01-01

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood

  8. Coupling of RO-MSF hybrid desalination plants with nuclear reactors

    International Nuclear Information System (INIS)

    Al-Sulaiman, Khalil; Al-Mutaz, Ibrahim S.

    1999-01-01

    Full text.Reverse osmosis (RO) and multistage flash (MSF) desalination are the most widely commercial available processes. MSF utilizes stream in the brine heater as a primary source of energy. RO is derived mainly by electricity that pumps the feed water against the mambranes. Steam and electricity and be produced easily by nuclear reactors. Nuclear reactors may be coupled with deslination plants (MSF, RO or combined (hybrid) RO/MSF configuration). This integrated plant will be capable of producing power and water at reasonable cost. The capital and operating cost will be reduced and the excess power can be efficiently utilized. Maintenance and operating cost will drop significantly. In this paper, a techno-economic study of hybrid reverses osmosis /multistage flash desalination will be carried. The proposed configuration (hybrid RO/MSF) coupled with nuclear reactor is considered the most appropriate candidate system for the application of dual-purpose nuclear desalination plants. the design parameters for such a desalination hybrid system will be the applied pressure and recovery for reverse osmosis plant and the number of stages and the heat transfer areas for multistage flash plant

  9. In-plant considerations for optimal offsite response to reactor accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Heising, C.D.; Aldrich, D.C.

    1982-11-01

    Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during accident situations. A simple plant information-management scheme is developed for use in offsite response decision-making. Detailed consequence calculations performed with the CRAC2 model are used to determine the appropriate timing of offsite-response implementation for a range of PWR accidents involving the reactor core. In-plant decision criteria for offsite-response implementation are defined. The definition of decision criteria is based on consideration of core-accident physical processes, in-plant accident monitoring information, and results of consequence calculations performed to determine the effectiveness of various public-protective measures. The benefits and negative aspects of the proposed response-implementation criteria are detailed

  10. N Reactor Production Assurance Program blance of plant evaluation: report of findings and conclusions

    International Nuclear Information System (INIS)

    Hurd, E.N.; Bitten, E.J.

    1985-03-01

    Fourteen tasks were identified by UNC Nuclear Industries for evaluating the life expectancy of N Reactor structures, systems and components in the Balance of Plant portion of the Production Assurance Program. A Westinghouse Hanford Company (WHC) evaluation team was assigned to each of these fourteen tasks. A uniform set of criteria was used by all teams in evaluating the problems that may be encountered during the extended plant operating lifetime. The overall conclusion is that extended life to those Balance of Plant components and systems studied can be achieved. Problems with the potential for compromising that conclusion are identified, and feasible solutions are provided

  11. The effect of uncertainties in nuclear reactor plant-specific failure data on core damage frequency

    International Nuclear Information System (INIS)

    Martz, H.F.

    1995-05-01

    It is sometimes the case in PRA applications that reported plant-specific failure data are, in fact, only estimates which are uncertain. Even for detailed plant-specific data, the reported exposure time or number of demands is often only an estimate of the actual exposure time or number of demands. Likewise the reported number of failure events or incidents is sometimes also uncertain because incident or malfunction reports may be ambiguous. In this report we determine the corresponding uncertainty in core damage frequency which can b attributed to such uncertainties in plant-specific data using a simple but typical nuclear power reactor example

  12. Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

    International Nuclear Information System (INIS)

    Davis, G.A.

    1992-01-01

    Since 1985, ABB Combustion Engineering Nuclear Power (CENP) and Duke Engineering ampersand Services, Inc. (DE ampersand S) have been developing the next generation of pressurized water reactor (PWR) plant for worldwide deployment. The goal is to make available a pre-licensed, standardized plant design that can satisfy the need for a reliable and economic supply of electricity for residential, commercial and industrial use. To ensure that such a design is available when needed, it must be based on proven technology and established licensing criteria. These requirements dictate development of nuclear technology that is advanced, yet evolutionary in nature. This has been achieved with the System 80+ Standard Plant Design

  13. Experimental investigations on the coolability of prototypical particle beds with respect to reactor safety; Experimentelle Untersuchungen der Kuehlbarkeit prototypischer Schuettungskonfigurationen unter dem Aspekt der Reaktorsicherheit

    Energy Technology Data Exchange (ETDEWEB)

    Leininger, Simon

    2017-02-22

    In case of a severe accident in a light water reactor, continuous unavailability of cooling water to the reactor core may result in overheating of the fuel elements and finally the loss of core integrity. Under such conditions, a structure of heat-releasing particles of different size and shape may be formed by fragmentation of molten core material in several stages of the accident. The long-term coolability of such beds is of prime im-portance to avoid any damage to the reactor pressure vessel or even a release of fission products to the environment. In the frame of this work, specific experiments were con-ducted under prototypical conditions employing the existing DEBRIS test facility in order to gain further knowledge about the thermohydraulic behavior of such beds. In steady state boiling experiments, the pressure gradients in particle beds were meas-ured both for one- and multi-dimensional cooling water flow conditions and compared with one another in order to assess the flow behavior inside the bed. For these different flow conditions as well as for stratified bed configurations, the maximum removable heat flux densities were determined in the dryout experiments. E. g., it was found that an axial stratification of the permeability can significantly reduce the bed's coolability. For the first time, the quenching behavior of dry, superheated beds was investigated at elevated system pressure up to 0.5 MPa. In these experiments, the effect of system pressure on the coolability was quantified by means of the quenching time (time period to cool down the bed to saturation temperature). The investigated particle beds mainly consisted of non-spherical particles with well-defined geometry (cylinders and screws). It was shown that the effect of the particles geometry on the flow in a particle bed can be best estimated by using an equivalent particle diameter calculated for monodisperse particle beds from the product of the Sauter diameter and a shape factor and for

  14. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  15. Nuclear power plant control room task analysis. Pilot study for pressurized water reactors

    International Nuclear Information System (INIS)

    Barks, D.B.; Kozinsky, E.J.; Eckel, S.

    1982-05-01

    The purposes of this nuclear plant task analysis pilot study: to demonstrate the use of task analysis techniques on selected abnormal or emergency operation events in a nuclear power plant; to evaluate the use of simulator data obtained from an automated Performance Measurement System to supplement and validate data obtained by traditional task analysis methods; and to demonstrate sample applications of task analysis data to address questions pertinent to nuclear power plant operational safety: control room layout, staffing and training requirements, operating procedures, interpersonal communications, and job performance aids. Five data sources were investigated to provide information for a task analysis. These sources were (1) written operating procedures (event-based); (2) interviews with subject matter experts (the control room operators); (3) videotapes of the control room operators (senior reactor operators and reactor operators) while responding to each event in a simulator; (4) walk-/talk-throughs conducted by control room operators for each event; and (5) simulator data from the PMS

  16. Civil ships propulsion reactor plants development and operation experience, and prospects for their improvement

    International Nuclear Information System (INIS)

    Vasyukov, V.I.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

    2000-01-01

    Russia is alone country in the World possessing nuclear-powered icebreaker fleet. Phases of creation in Russia of several propulsion nuclear reactor plant generations are considered. By present 8 nuclear ice-breakers and a nuclear-powered cargo ship (lighter carrier) have been constructed, for which three of propulsion NSSS generations were developed. Their brief description, main performance indicators and results of operation since 1959 are given. It is shown that gradual evolution of NSSS design features has ensured creation of reliable, safe and environmentally friendly propulsion reactor plants. Issues of the propulsion NSSS life extension and improvement for new generation of nuclear ice-breakers, cargo ships, floating heat and power plants, sea water desalination and power generating complexes are considered with account for the gained operating experience. Activities on creation of new generation nuclear-powered ships and floating NPPs are considered as prospective sphere of Russia's collaboration with other countries of the World community. (author)

  17. Aging management strategy for reactor internals of Korean nuclear power plants

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

    2010-01-01

    This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

  18. A dynamic model of the reactor coolant system flow for KMRR plant simulation

    International Nuclear Information System (INIS)

    Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

    1990-01-01

    To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

  19. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Kuruc, J.; Matel, L.

    2007-01-01

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

  20. Potential environmental effects of fusion reactor power plants

    International Nuclear Information System (INIS)

    Young, J.R.; Gore, B.F.; Coffman, F.E.

    1976-01-01

    Construction and operation of fusion power plants is expected to reduce the total environmental effects of 21st century power generation. Fusion power plant impacts due to noise, odors, vibrations, and sanitary wastes are expected to be insignificant. impacts due to land use, chemical releases, and aesthetics are expected to be reduced. Impacts due to heat releases, local socio-economic changes, and non-radioactive liquid and solid disposal are expected to be comparable to those for the alternative fission or coal-fired power systems. Radiation doses to the public due to radioactive wastes are expected to be comparable to, or less than, the trivial low doses due to fission power systems. Research and development will be required, however, to assure adequate containment of tritium, the primary radioisotope of concern. Prevention of accidental tritium releases is within the capability of current engineering practice. Current technology is capable of handling the solid radioactive waste which may be produced, with insignificant environmental impact. Major research efforts are necessary to determine if subtle long-term effects of magnetic fields exist and should be of concern. In view of the large quantities of construction materials required for fusion. Material availability may dictate 21st century power plant design and construction. The accident potential of fusion power plants should be lower than for fission systems. Accidental criticalities and plasma runaways are not considered to be possible. Loss of coolant accidents are not expected to result in damage to the containment. No fission products or actinides are present to be released in an accident, and most activation products are immobilized in structures. The biological hazard of tritium is orders of magnitude smaller than for fission products and actinides. Safeguards against diversion of fissile materials are not expected to be necessary

  1. Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1989-01-01

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. This paper reviews the research and development (R and D) programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

  2. Considerations on safety against seismic excitations in the project of reactor auxiliary building and control building in nuclear power plants

    International Nuclear Information System (INIS)

    Santos, S.H.C.; Castro Monteiro, I. de

    1986-01-01

    The seismic requests to be considered in the project of main buildings of a nuclear power plant are discussed. The models for global seismic analysis of nuclear power plant structures, as well as models for global strength distribution are presented. The models for analysing reactor auxiliary building and control building, which together with the reactor building and turbine building form the main energy generation complex in a nuclear power plant, are described. (M.C.K.) [pt

  3. Multi-stage-flash desalination plants of relative small performance with integrated pressurized water reactors as a nuclear heat source

    International Nuclear Information System (INIS)

    Petersen, G.; Peltzer, M.

    1977-01-01

    In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination plants with a performance in the range of 10 000 to 80 000 m 3 distillate per day heated by a nuclear reactor are investigated. The reactor concept is similar to the Integrated Pressurized Water Reactor (IPWR) of the nuclear ship OTTO HAHN. The design study shows that IPWR systems have specific advantages up to 200 MWth compared to other reactor types at least being adapted for single- and dual-purpose desalination plants. The calculated costs of the desalinated water show that due to fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (author)

  4. Focused proton beams propagating in reactor of fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Niu, K [Teikyo Heisei Univ., Uruido, Ichihara, Chiba (Japan)

    1997-12-31

    One of the difficult tasks of light ion beam fusion is to propagate the beam in the reactor cavity and to focus the beam on the target. The light ion beam has a certain local divergence angle because there are several causes for divergence at the diode. The electrostatic force induced at the leading edge causes beam divergence during propagation. To confine the beam within a small radius during propagation, the magnetic field must be employed. Here the electron beam is proposed to be launched simultaneously with the launching of the proton beam. If the electron beam has the excess current, the beam induces a magnetic field in the negative azimuthal direction, which confines the ion beam within a small radius by the electrostatic field as well as the electron beam by the Lorentz force. The metal guide around the beam path helps the beam confinement and reduces the total amount of magnetic field energy induced by the electron current. (author). 2 figs., 15 refs.

  5. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  6. Recent advances in AFB biomass gasification pilot plant with catalytic reactors in a downstream slip flow

    Energy Technology Data Exchange (ETDEWEB)

    Aznar, M P; Gil, J; Martin, J A; Frances, E; Olivares, A; Caballero, M A; Perez, P [Saragossa Univ. (Spain). Dept. of Chemistry and Environment; Corella, J [Madrid Univ. (Spain)

    1997-12-31

    A new 3rd generation pilot plant is being used for hot catalytic raw gas cleaning. It is based on a 15 cm. i.d. fluidized bed with biomass throughputs of 400-650 kg/h.m{sup 2}. Gasification is performed using mixtures of steam and oxygen. The produced gas is passed in a slip flow by two reactors in series containing a calcined dolomite and a commercial reforming catalyst. Tars are periodically sampled and analysed after the three reactors. Tar conversions of 99.99 % and a 300 % increase of the hydrogen content in the gas are obtained. (author) (2 refs.)

  7. Recent advances in AFB biomass gasification pilot plant with catalytic reactors in a downstream slip flow

    Energy Technology Data Exchange (ETDEWEB)

    Aznar, M.P.; Gil, J.; Martin, J.A.; Frances, E.; Olivares, A.; Caballero, M.A.; Perez, P. [Saragossa Univ. (Spain). Dept. of Chemistry and Environment; Corella, J. [Madrid Univ. (Spain)

    1996-12-31

    A new 3rd generation pilot plant is being used for hot catalytic raw gas cleaning. It is based on a 15 cm. i.d. fluidized bed with biomass throughputs of 400-650 kg/h.m{sup 2}. Gasification is performed using mixtures of steam and oxygen. The produced gas is passed in a slip flow by two reactors in series containing a calcined dolomite and a commercial reforming catalyst. Tars are periodically sampled and analysed after the three reactors. Tar conversions of 99.99 % and a 300 % increase of the hydrogen content in the gas are obtained. (author) (2 refs.)

  8. Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Houtman, J.L.

    1979-01-01

    The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

  9. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  10. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  11. A small floating seawater desalination plant using a nuclear heating reactor coupled with the MED process

    International Nuclear Information System (INIS)

    Dong Duo; Wu Shaorong; Zhang Dafang; Wu Zongxin

    1997-01-01

    A small floating seawater desalination plant using a nuclear heating reactor coupled with a multi-effect distillation (MED) process was designed by the Institute of Nuclear Energy Technology, Tsinghua University of China. It was intended to supply potable water to remove coastal areas or islands where both fresh water and energy are severely lacking, and also to serve as a demonstration and training facility. The design of a small floating plant coupled two proven technologies in the cogeneration mode: a nuclear heating reactor (NHR-10), with inherent, passive safety features based on NHR-5 experience, and a low temperature MED process. The secondary loop was designed as a safety barrier between the primary loop and the steam loop. With a 10 MW(th) heating reactor, the floating plant could provide 4,000 m 3 /d of potable water and 750 kW of electricity. The design concept and parameters, safety features, coupling scheme and floating plant layout are presented in the paper. (author). 3 refs, 4 figs, 3 tabs

  12. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  13. Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

    1999-05-15

    The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

  14. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  15. Nonlinear Dynamic Model of Power Plants with Single-Phase Coolant Reactors

    International Nuclear Information System (INIS)

    Vollmer, H.

    1968-12-01

    The traditional way of developing dynamic models for a specific nuclear power plant and for specific purpose seems rather uneconomical, as much of the information often can not be utilized if the plant design or the required accuracy of the calculation is desired to be changed. It is therefore suggested that the model development may be made more systematic, general and flexible by - applying the 'box of bricks' system, where the main components of a nuclear power plant are treated separately and combined afterwards according to a given flow scheme, - a dynamic determination of the components which is as general as possible without taking into account those details which have a minor influence on the overall dynamics, - providing approximations of the more rigorous solution sufficient to meet the user s requirements on accuracy, - proper use of computers. A dynamic model for single-phase coolant reactor plants is established along these lines. By separation of the nonlinear and linear parts of the system, application of Laplace transformation and proper approximations, and the use of a hybrid computer it seems possible to determine the (nonlinear) dynamic behaviour of such a plant for perturbations which are not so large that phase changes of physical parameters occur, e. g. fuel does not melt. The model is applied to a steam cooled fast reactor power plant

  16. Development of graphic display program of reactor operating parameters for emergency exercise at nuclear power plants

    International Nuclear Information System (INIS)

    Okuda, Yasunori; Yoshida, Yoshitaka; Gotou, Kazuko

    2001-01-01

    A scenario of nuclear emergency exercise based on the result of accident progress analysis is expected to ensure effective training. Thereupon a new graphic display program for reactor operating parameters has been developed to present real-time of plant process values (parameters), released radioactivities from the plant, and dose rate data around the site calculated by using the accident analysis code MAAP4 and other codes. This system has a trend graph screen displaying reactor operating parameters, an environmental dose rate summary screen indicating dose rate distribution around the site on the map, and a plant parameters summary screen showing important plant parameters on a simplified plant system diagram. One screen can be switched to another any time. It also has a jump-function easily accessing any stage during the exercise scenario in accordance with progress of the exercise. As a result of the application of this system to a real nuclear emergency exercise, it has been verified that this system is quite useful for confirming the parameters when the nuclear emergency exercise starts and the licensee reports the plant conditions to related bodied. (author)

  17. Development of graphic display program of reactor operating parameters for emergency exercise at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Okuda, Yasunori; Yoshida, Yoshitaka [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Gotou, Kazuko [Kansai Electric Power Co., Inc., Osaka (Japan)

    2001-09-01

    A scenario of nuclear emergency exercise based on the result of accident progress analysis is expected to ensure effective training. Thereupon a new graphic display program for reactor operating parameters has been developed to present real-time of plant process values (parameters), released radioactivities from the plant, and dose rate data around the site calculated by using the accident analysis code MAAP4 and other codes. This system has a trend graph screen displaying reactor operating parameters, an environmental dose rate summary screen indicating dose rate distribution around the site on the map, and a plant parameters summary screen showing important plant parameters on a simplified plant system diagram. One screen can be switched to another any time. It also has a jump-function easily accessing any stage during the exercise scenario in accordance with progress of the exercise. As a result of the application of this system to a real nuclear emergency exercise, it has been verified that this system is quite useful for confirming the parameters when the nuclear emergency exercise starts and the licensee reports the plant conditions to related bodied. (author)

  18. Nonlinear Dynamic Model of Power Plants with Single-Phase Coolant Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollmer, H

    1968-12-15

    The traditional way of developing dynamic models for a specific nuclear power plant and for specific purpose seems rather uneconomical, as much of the information often can not be utilized if the plant design or the required accuracy of the calculation is desired to be changed. It is therefore suggested that the model development may be made more systematic, general and flexible by - applying the 'box of bricks' system, where the main components of a nuclear power plant are treated separately and combined afterwards according to a given flow scheme, - a dynamic determination of the components which is as general as possible without taking into account those details which have a minor influence on the overall dynamics, - providing approximations of the more rigorous solution sufficient to meet the user s requirements on accuracy, - proper use of computers. A dynamic model for single-phase coolant reactor plants is established along these lines. By separation of the nonlinear and linear parts of the system, application of Laplace transformation and proper approximations, and the use of a hybrid computer it seems possible to determine the (nonlinear) dynamic behaviour of such a plant for perturbations which are not so large that phase changes of physical parameters occur, e. g. fuel does not melt. The model is applied to a steam cooled fast reactor power plant.

  19. 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.)

  20. Manipulator and materials handling systems for reactor decommissioning -Cooperation between the university and the plant operator

    International Nuclear Information System (INIS)

    Schreck, G.; Bach, F. W.; Haferkamp, H.

    1995-01-01

    Nuclear reactor dismantling requires suitable handling systems for tools and disassembled components, as well as qualified and reliable disassembly and cutting techniques. From the angle of radiation protection, remote-controlled handling techniques and underwater techniques are the methods of choice, the latter particularly in continuation of plant operating conditions, and this all the more the more disassembly work proceeds towards the reactor core. With the experience accumulated for 20 years now by the Institut fuer Werkstoffkunde (materials science) of Hannover University by basic research and application-oriented development work in the field of thermal cutting technology, especially plasma arc cutting techniques, as well as development work in the field of remote-controlled materials handling systems, the institute is the cut-out partner for disassembly tasks in reactor decommissioning. (Orig./DG) [de

  1. BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Hetrick, D.L.; Sowers, G.W.

    1978-06-01

    This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices

  2. Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors

    International Nuclear Information System (INIS)

    1985-07-01

    This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55

  3. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

  4. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  5. Project of SVBR-75/100 reactor plant with improved safety for nuclear sources of small and medium power

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Stepanov, V. S.; Klimov, N. N.; Dedul, A. V.; Bolvanchikov, S. N.; Zrodnikov, A. V.; Tolhinsky, G. I.; Komlev, O. G.

    2004-01-01

    As a result of the joint work performed recently by FSUE OKB Gidropress, SNC RF-IPPE and other organizations the technical feasibility is shown for creation and usage in nuclear power engineering of the unified reactor plant (RP) SVBR-75/100 with fast neutron reactor core and lead-bismuth coolant (LBC) in the primary circuit. Technical design of SVBR-75/100 reactor plant is based on the following: 50-year operation experience in development and operation of RP with LBC for nuclear submarines; experience in development and operation of fast reactor with sodium coolant; experience in optimization of LBC technology at nuclear submarines and ground-based test benches; conceptual design of SVBR-75 reactor plant (for renovation of Units 2, 3 and 4 of Novovoronezh NPP). Technical solutions laid down in the basis of SVBR-75/100 reactor plant design are oriented towards the industrial basis, structural materials existing in Russia, as well as the unique LBC technology with experimental and practical support. The concept of SVBR-75/100 reactor plant safety assurance is based on the following provisions: maximum usage of inherent safety supported by physical features of fast neutron reactor, chemically inert LBC in the primary circuit, integral layout and special design solutions; maximum possible combination of normal operation and safety functions in RP systems. Small power of SVBR-75/100 RP makes it possible to manufacture the complete set of RP main equipment at the factory and delivery it to NPP site as-finished practically using any transport including railway. Possible fields of application of SVBR-75/100 reactor plants: modular NPPs of different power; renovation of NPPs with light water reactors exhausted their service life; independent nuclear power sources for different applications (ground-based nuclear water-desalinating plants, etc. )(author)

  6. MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR

    Directory of Open Access Journals (Sweden)

    Suwardi Suwardi

    2015-03-01

    Full Text Available MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*. The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m. MODEL SIMULASI VARIASI GEOMETRI DAN STRESS-STRAIN DARI PROTOTIP BAHAN BAKAR PIN BATAN SELAMA UJI IRADIASI DI REAKTOR RSG-GAS. Pusat Teknologi Bahan Bakar Nuklir (PTBBN telah menyiapkan tangkai (pin bahan bakar pendek perdana yang berisi pelet UO2 alam dalam kelongsong paduan zircaloy untuk dilakukan uji iradiasi daya naik. Penelitian ini merupakan bagian dari perancangan percobaan iradiasi pertama di PRTF (Power Ramp Test Fasility yang terpasang di reaktor serbaguna RSG-GAS berdaya 30 MW. Telah dilakukan pemodelan dan simulasi kinerja termal mekanikal pin selama iradiasi. Variasi geometri pelet dan kelongsong selama pengujian dimodelkan dengan memperhatikan fenomena ekspansi termal

  7. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  8. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto

    2017-01-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10"1"3 combinations and 10"1"1 great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  9. Near term feasibility of nuclear reactor for sea-water desalting: coupling of standard condensing nuclear power stations to low grade heat multieffect distillation plants

    International Nuclear Information System (INIS)

    Adar, J.; Manor, S.; Schaal, M.

    1977-01-01

    Commercial nuclear power reactors exist only in standard sizes and designs. No large nuclear back-pressure turbines are available today. Therefore, near term large scale nuclear desalination plants must be tailored to the NSSS sizes and available turbines and not the contrary. Standard condensing nuclear turbines could operate continuously with a back-presure of up to 5-7'' Hg (depending on the supplier). It means that they can exhaust huge amounts of steam at 56 0 C - 64 0 C with a loss of electricity production of 6% - 10% when compared to 2 1/2'' Hg normal condensing pressure. The horizontal aluminium tube multi-effect distillation process developed by ''Israel Desalination Engineering'' Ltd. is very suitable for the use of such low-grade heat: 4 to 9 effects can operate within these temperature ranges. A special flash-chamber constitutes a positive barrier against any possible contamination being carried over by the steam exhausted from the turbine to the desalination plant. Flow sheets, heat and mass balances have been prepared for two standard sizes of NSSS and turbines (1882sup(Mwth) and 2785sup(Mwth)), two ''back-pressures'' (5 1/2'' and 7'' Hg), and corresponding desalination plants. Only standard equipment is being used in the steam and electricity producing plant. The desalination plant consists of 6 to 12 parallel double lines, each of them similar to a large prototype now being designed and which is going to be coupled to an old fossil power station. Water production varies between 50 and 123 sup(us MGD) and water cost between 23 and 36 sup(cents)/M 3 . Total energy requirements of the desalination plant represent only 19 to 50% of the total water cost as against 75% for a single purpose plant. Costs are based on actual bids for the power plant and actual estimates for the desalination prototype. The operation is designed to be flexible so that the power plant can be operated either in conjunction with the desalination plant, or as a single purpose

  10. 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

  11. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  12. An overview of reactor vessel internals segmentation for nuclear plant decommissioning

    International Nuclear Information System (INIS)

    Litka, T.J.

    1994-01-01

    Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

  13. A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

    1987-01-01

    A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

  14. At-reactor storage of spent fuel for life-of-plant

    International Nuclear Information System (INIS)

    Fuierer, A.A.

    1990-01-01

    The management of commercial spent fuel is a fairly broad topic beginning with the discharge from a reactor, its storage on-site, its transport from the reactor site to a U.S. Department of Energy (DOE) facility, and its ultimate disposal in a geologic repository. This paper discusses spent-fuel management in the at-reactor phase. There are two basic methods for at-reactor storage of spent fuel. The first is wet storage in a pool, and the second is dry storage external to the plant in some form of cask or vault. Spent-fuel consolidation will impact the utility and the DOE waste system. Some of these impacts have a positive effect and some have a negative effect, and each will vary somewhat for each utility. Spent-fuel consolidation and life-of-plant storage will be an increased burden to utilities but will likely result in significant cost savings to the overall waste management system and by proper integration can result in significant institutional benefits

  15. A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

    International Nuclear Information System (INIS)

    Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

    2017-01-01

    This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

  16. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    International Nuclear Information System (INIS)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L.

    2015-09-01

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  17. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    Energy Technology Data Exchange (ETDEWEB)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2015-09-15

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  18. Prototyping Practice

    DEFF Research Database (Denmark)

    Ramsgaard Thomsen, Mette; Tamke, Martin

    2015-01-01

    This paper examines the role of the prototyping in digital architecture. During the past decade, a new research field has emerged exploring the digital technology’s impact on the way we think, design and build our environment. In this practice the prototype, the pavilion, installation or demonstr......This paper examines the role of the prototyping in digital architecture. During the past decade, a new research field has emerged exploring the digital technology’s impact on the way we think, design and build our environment. In this practice the prototype, the pavilion, installation...

  19. Dynamic simulation of a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Shinaishin, M.A.M.

    1976-01-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is dealt with. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. In addition to the usual assumption of lumped parameters, uniform heat transfer and point kinetics (prompt jump) have been the main approximations in this and other simulators (see below). Two different transport-delay models have also been installed in all simulators. The simulators were constructed using the DARE-P System, developed by the Electrical Engineering Department at the University of Arizona

  20. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.