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Sample records for dismantling reactor pressure

  1. Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

    International Nuclear Information System (INIS)

    Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

    2008-01-01

    The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

  2. Dismantling method for reactor pressure vessel and system therefor

    International Nuclear Information System (INIS)

    Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

    1994-01-01

    Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

  3. International Cooperation for the Dismantling of Chooz A Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Grenouillet, J.J.; Posivak, E.

    2009-01-01

    Chooz A is the first PWR that is being decommissioned in France. The main issue that is conditioning the success of the project is the Reactor Pressure Vessel (RPV) and Reactor Vessel Internals (RVI) segmentation. Whereas Chooz A is the first and unique RPV and RVI being dismantled in France, there are many similar experiences available in the world. Thus the project team was eager to cooperate with other teams facing or being faced with the same issue. A cooperation programme was established in two separate ways: - Benefiting from experience feedback from completed RPV and RVI dismantling projects, - Looking for synergy with future RPV dismantling projects for activities such as segmentation tools design, qualification and manufacturing for example. This paper describes the implementation of this programme and how the outcome of the cooperation was used for the implementation of Chooz-A RPV and RVI segmentation project. It shows also the limits of such a cooperation. (authors)

  4. Underwater cutting of stainless steel plate and pipe for dismantling reactor pressure vessels

    International Nuclear Information System (INIS)

    Hamasaki, M.; Tateiwa, F.; Kanatani, F.; Yamashita, S.

    1982-01-01

    A consumable electrode water jet cutting technique is described. Satisfactory underwater cutting of 80mm stainless steel plate using a current of 2000A and at a water depth of 200mm has been demonstrated. The electrical requirements for this arc welding method applied to cutting were found to be approximately one third those required for conventional plasma arc cutting for the same thickness plate. An application of this technique might be found in the dismantling of atomic reactor pressure vessels, and parts of commercial atomic reactors. (author)

  5. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    Energy Technology Data Exchange (ETDEWEB)

    Boucau, Joseph [Westinghouse Electric Company, 43 rue de l' Industrie, Nivelles (Belgium); Mirabella, C. [Westinghouse Electric France, Orsay (France); Nilsson, Lennart [Westinghouse Electric Sweden, Vaesteraas (Sweden); Kreitman, Paul J. [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States); Obert, Estelle [EDF - DPI - CIDEN, Lyon (France)

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French

  6. The timing of reactor dismantling

    International Nuclear Information System (INIS)

    Roberts, P.

    2000-01-01

    Work has been progressing across the world for the decommissioning of nuclear reactors. The initial work focused on the early, complete dismantling but this was associated with small size reactors and was done for experimental or demonstration purposes. The situation now is that an increasing number of full size power reactors are being shutdown and decision are being made as to the decommissioning strategy to be applied, e.g. with respect to the appropriate timing of reactor dismantling. There are two basic approaches to the timing of reactor dismantling, which are to either proceed with dismantling on an early time scale or to delay it for a period of years. There are a number of examples worldwide of both approaches being taken but one common feature of the approach taken by most countries is that decisions are made on a case by case basis, taking account of relevant factors, and as a result the strategy can vary from reactor to reactor and from country to country. Decisions on timing take account of the following main factors: safety, radioactive decay, financial factors, radioactive waste, reactor type, technology, repository availability, site re-use, regulatory standards, plant knowledge/records, other issues

  7. EL-3 dismantling of an experimental reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

  8. The dismantling of fast reactors: sodium processing

    International Nuclear Information System (INIS)

    Rodriguez, G.; Berte, M.; Serpante, J.P.

    1999-01-01

    Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

  9. Decommissioning and dismantling reactors and managing waste

    International Nuclear Information System (INIS)

    Bensoussan, E.; Reicher-Fournel, N.

    2005-01-01

    In the early forties/fifties, a number of countries launched the first developments in the field of nuclear power. Some of them now have large numbers of nuclear facilities and nuclear power plants which have met, and continue to meet, the objectives for which they were designed and built. Other plants, including nuclear fuel production and enrichment plants, experimental reactors or research reactors, will have to be dismantled and demolished in the near future. These activities are handled differently in different countries as a function of specific energy policies, advanced development plants, current financial resources, the availability of qualified engineers and specialized industries able to handle projects of this kind, as well as other factors. All dismantling and demolition projects serve the purpose of returning the respective sites to green-field conditions. (orig.)

  10. Chooz A: a model for the dismantling of water-cooled reactors

    International Nuclear Information System (INIS)

    Anon.

    2017-01-01

    The specificity of Chooz-A, the first French pressurized water reactor (PWR), is that the reactor and its major components (pumps, exchangers and cooling circuits) are installed in 2 caves dug out in a hill slope. Chooz-A was operating from 1967 to 1991, in 1993 the fuel was removed and in 2007 EDF received the authorization to dismantle the reactor. In 2012, EDF completed the dismantling of the cave containing the elements of the cooling circuit, a cornerstone was the removing of the four 14 m high steam generators. The dismantling of the pressure vessel began in march 2017, it is the same tools and the same processes that were used for the dismantling of the pressure vessel of the Zorita plant (Spain) in 2016. The end of the Chooz-A dismantling is expected in 2022. The feedback experience will help to standardize practices for the French fleet of PWRs. (A.C.)

  11. Dismantling the nuclear research reactor Thetis

    Energy Technology Data Exchange (ETDEWEB)

    Michiels, P. [Belgoprocess, 2480 Dessel (Belgium)

    2013-07-01

    The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were

  12. The Japan Power Demonstration Reactor (JPDR) dismantling activities. Management of JPDR dismantling waste

    International Nuclear Information System (INIS)

    Abe, Masayoshi; Nakata, Susumu; Ito, Shinichi

    1996-01-01

    The management of wastes, both radioactive and non-radioactive, is one of the most important issues for a safe and reasonable dismantling operation of nuclear power plants. A large amount of radioactive wastes is arising from a reactor dismantling operation in a relatively short period time, ranging in a wide variety from very low level to relatively high level. Moreover non-radioactive waste is also in a huge amount. The dismantling operation of Japan Power Demonstration Reactor (JPDR) resulted in 24,440 tons of dismantling wastes, of which about 15% was radioactive and 85% non-radioactive. These wastes were managed successfully implementing a well developed management plan for JPDR dismantling waste. Research and development works for handling of JPDR dismantling wastes were performed, including fixation of loose contamination on surface, volume reduction and waste containers for on-site transportation and interim storage. The JPDR dismantling wastes generated were classified and categorized depending on their materials, characteristics and activity level. Approximately 2,100 tons of radioactive wastes were stored in the interim storage facilities on site using developed containers, and 1,670 tons of radioactive concrete waste were used for a safe demonstration test of a simple near-surface disposal for very low level waste. Other dismantling wastes such as steel and concrete which were categorized as non-radioactive were recycled and reused as useful resources. This paper describes the management of the JPDR dismantling wastes. (author)

  13. Method of freezing type dismantling for wasted reactors

    International Nuclear Information System (INIS)

    Tatsumi, Toshiyuki.

    1985-01-01

    Purpose: To enable to operate a cutting device in the air by placing a working table on ice while utilizing the ice as radiation shielding materials thereby prevent the diffusion of air contaminations. Method: Upon dismantling a BWR type reactor, ice is packed into a reactor container and a pressure vessel and frozen state is maintained by cooling coils disposed to the outer circumference of the pressure vessel. Then, an airtight hood is covered over the pressure vessel and a working table is rotatably disposed therein. Upon working, when the upper layer ice is melted by a heat pump and discharged, the airtight hood is lowered to a predetermined level. After freezing the melted portion again at the lowered level, cutting work is conducted by an operator in the hood. The cut pieces are conveyed after hoisting the airtight hood by a crane. The pressure vessel is dismantled by repeating the foregoing procedures. In this way, cut pieces can be recovered without falling them to the reactor bottom as in the conventional work in water. In addition, since the procedures are conducted while covering the airtight hood, diffusion of air contaminations can be prevented. (Kamimura, M.)

  14. Method of dismantling a nuclear reactor

    International Nuclear Information System (INIS)

    Shirai, Masato; Hashimoto, Osamu.

    1984-01-01

    Purpose: To enable rapid and simple positioning for a plasma arc torch disposed to the inside of a nuclear reactor main body. Method: After removing the upper semi-spherical portion, fuel portion and control rod portion of a nuclear reactor, a rotary type girder is placed on the upper edge of a cylindrical portion remained after the removal of the upper semi-spherical portion. Then, the upper portion of a supporting rod provided with a swing arm having a plasma arc torch at the top end is situated at the center of the reactor main body. Then, the top end of the support rod is inserted to fix in the housing of control rod drives. Then, the swing arm is actuated to situate the plasma arc torch to a desired position to be cut, whereafter cutting is initiated while rotating the rotary type girder. Thus, plasma arc torch is moved horizontally along an arcuate trace, whereby pipeways, accessories or the likes disposed to the inside of the main body are at first cut and then the cylindrical portion constituting the main body is cut to dismantle the reactor. (Moriyama, K.)

  15. Advanced robotic remote handling system for reactor dismantlement

    International Nuclear Information System (INIS)

    Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

    1991-01-01

    An advanced robotic remote handling system equipped with a multi-functional amphibious manipulator has been developed and used to dismantle a portion of radioactive reactor internals of an experimental boiling water reactor in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. (author)

  16. Decommissioning of the AVR reactor, concept for the total dismantling

    International Nuclear Information System (INIS)

    Marnet, C.; Wimmers, M.; Birkhold, U.

    1998-01-01

    After more than 21 years of operation, the 15 MWe AVR experimental nuclear power plant with pebble bed high temperature gas-cooled reactor was shout down in 1988. Safestore decommissioning began in 1994. In order to completely dismantle the plant, a concept for Continued dismantling was developed according to which the plant could be dismantled in a step-wise procedure. After each step, there is the possibility to transform the plant into a new state of safe enclosure. The continued dismantling comprises three further steps following Safestore decommissioning: 1. Dismantling the reactor vessels with internals; 2. Dismantling the containment and the auxiliary units; 3. Gauging the buildings to radiation limit, release from the validity range of the AtG (Nuclear Act), and demolition of the buildings. For these steps, various technical procedures and concepts were developed, resulting in a reference concept in which the containment will essentially remain intact (in-situ concept). Over the top of the outer reactor vessel a disassembling area for remotely controlled tools will be erected that tightens on that vessel and can move down on the vessel according to the dismantling progress. (author)

  17. Development of multi-functional telerobotic systems for reactor dismantlement

    International Nuclear Information System (INIS)

    Fujii, Yoshio; Usui, Hozumi; Shinohara, Yoshikuni

    1992-01-01

    This report summarizes technological features of advanced telerobotic systems for reactor dismantling application developed at the Japan Atomic Energy Research Institute. Taking into consideration the special environmental conditions in reactor dismantling, major effort was made to develop multifunctional telerobotic system of high reliability which can be used to perform various complex tasks in an unstructured environment and operated in an easy and flexible manner. The system development was carried out through constructing three systems in seccession; a light-duty and a heavy-duty system as a prototype system for engineering test in cold environment, and a demonstration system for practical on-site application to dismantling highly radioactive reactor internals of an experimental boiling water reactor JPDR (Japan Power Demonstration Reactor). Each system was equipped with one or two amphibious manipulators which can be operated in either a push-button manual, a bilateral master-slave, a teach-and-playback or a programmed control mode. Different scheme was adopted in each system at designing the manipulator, transporter and man-machine interface so as to compare their advantages and disadvantages. According to the JPDR decommissioning program, the demonstration system was successfully operated to dismantle a portion of the radioactive reactor internals of the JPDR, which used underwater plasma arc cutting method and proved the usefulness of the multi-functional telerobotic system for reducing the occupational hazards and enhancing the work efficiency in the course of dismantling highly radioactive reactor components. (author)

  18. Development of telerobotic manipulators for reactor dismantling work

    International Nuclear Information System (INIS)

    Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

    1991-01-01

    This paper describes the amphibious electrical manipulators JARM-10, JART-25, JART-100 and JARM-25 which were developed in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. They are multi-functional telerobotic light-duty (10 and 25 daN) and heavy-duty (100 daN) Manipulators which can be used in hostile environments in reactor dismantling work such as high radiation, underwater work and electrical noise. Each manipulator can be operated in either a bilateral master-slave, a teach-and-playback or a programmed control mode. By combining these modes appropriately, it is possible to perform complex tasks of remote handling. The usefulness of the telerobotic systems for dismantling nuclear reactors has been demonstrated by successful application of the JARM-25 for remote underwater dismantlement of highly radioactive reactor internals of complex form of an experimental nuclear power reactor. (author)

  19. The Japan Power Demonstration Reactor dismantling project. Radiation control

    International Nuclear Information System (INIS)

    Tomii, Hiroyuki; Seiki, Yoshihiro

    1996-01-01

    In the Japan Power Demonstration Reactor (JPDR) dismantling project, radiation control was performed properly with routine and special monitoring to keep the occupational safety and to collect data necessary for future dismantling of nuclear facilities. This report describes a summary of radiation control in the dismantling activities and some results of parametric analysis on dose equivalent evaluation, and introduces the following knowledge on radiological protection effectiveness of the dismantling systems applied in the project. a) Use of remote dismantling systems was effective in reducing equivalent workplace exposure. b) Utilization of existing facilities as radiation shield or radioactivity containment was effective in reducing workplace exposure, and also in increasing work efficiency. c) Use of underwater cutting systems was useful to minimize air contamination, and to reduce the dose equivalent rate in the working area. d) In the planning of dismantling, it is necessary to optimize the radiation protection by analyzing dismantling work procedures and evaluating radiological features of the dismantling systems applied, including additional work which the systems require brought from such activities. (author)

  20. Experience in dismantling and packaging of pressure vessel and core internals

    International Nuclear Information System (INIS)

    Pillokat, Peter; Bruhn, Jan Hendrik

    2011-01-01

    Nuclear Company AREVA is proud to look back on versatile experience in successfully dismantling nuclear components. After performing several minor dismantling projects and studies for nuclear power plants, AREVA completed the order for dismantling of all remaining Reactor Pressure Vessel internals at German Boiling Water Reactor Wuergassen NPP in October '08. During the onsite activities about 121 tons of steel were successfully cut and packed under water into 200l- drums, as the dismantling was performed partly in situ and partly in an underwater working tank. AREVA deployed a variety of different cutting techniques such as band sawing, milling, nibbling, compass sawing and water jet cutting throughout this project. After successfully finishing this task, AREVA dismantled the cylindrical part of the Wuergassen Pressure Vessel. During this project approximately 320 tons of steel were cut and packaged for final disposal, as dismantling was mainly performed by on air use of water jet cutting with vacuum suction of abrasive and kerfs material. The main clue during this assignment was the logistic challenge to handle and convey cut pieces from the pressure vessel to the packing area. For this, an elevator was installed to transport cut segments into the turbine hall, where a special housing was built for final storage conditioning. At the beginning of 2007, another complex dismantling project of great importance was acquired by AREVA. The contract included dismantling and conditioning for final storage of the complete RPV Internals of the German Pressurized Water Reactor Stade NPP. Very similar cutting techniques turned out to be the proper policy to cope this task. On-site activities took place in up to 5 separate working areas including areas for post segmentation and packaging to perform optimized parallel activities. All together about 85 tons of Core Internals were successfully dismantled at Stade NPP until September '09. To accomplish the best possible on

  1. Dismantling method for reactor shielding wall and device therefor

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1995-01-01

    A ring member having an outer diameter slightly smaller than an inner diameter of a reactor shielding wall to be dismantled is lowered in the inside of the reactor shielding wall while keeping a horizontal posture. A cutting device is disposed at the lower peripheral edge of the ring member. The cutting device can move along the peripheral edge of the circular shape of the ring member. The ring member is urged against the inner surface of the reactor shielding wall by using an urging member to immobilize the ring member. Then, the cutting device is operated to cut the reactor shielding wall into a plurality of ring-like blocks at a plurality of inner horizontal ribs or block connection ribs. Then, the blocks of the cut reactor shielding wall are supported by the ring member, and transported out of the reactor container by a lift. The cut blocks transported to the outside are finely dismantled for every block in a closed chamber. (I.N.)

  2. A Study on Dismantling of Westinghouse Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Lee, Sang-Guk

    2014-01-01

    KHNP started a research project this year to develop a methodology to dismantle nuclear reactors and internals. In this paper, we reviewed 3D design model of the reactor and suggested feasible cutting scheme.. Using 3-D CAD model of Westinghouse type nuclear reactor and its internals, we reviewed possible options for disposal. Among various options of dismantling the nuclear reactor, plasma cutting was selected to be the best feasible and economical method. The upper internals could be segmented by using a band saw. It is relatively fast, and easily maintained. For cutting the lower internals, plasma torch was chosen to be the best efficient tool. Disassembling the baffle and the former plate by removing the baffle former bolts was also recommended for minimizing storage volume. When using plasma torch for cutting the reactor vessel and its internal, installation of a ventilation system for preventing pollution of atmosphere was recommended. For minimizing radiation exposure during the cutting operation, remotely controlled robotic tool was recommended to be used

  3. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    Tas, A.; Teunissen, G.

    1986-01-01

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

  4. Decontamination before dismantling a fast breeder reactor primary cooling system

    International Nuclear Information System (INIS)

    Costes, J.R.; Antoine, P.; Gauchon, J.P.

    1997-01-01

    The large-scale decontamination of FBR sodium loops is a novel task, as only a limited number of laboratory-scale results are available to date. The principal objective of this work is to develop a suitable decontamination procedure for application to the primary loops of the RAPSODIE fast breeder reactor as part of decommissioning to Stage 2. After disconnecting the piping from the main vessel, the pipes were treated by circulating chemical solutions and the vessels by spraying. The dose rate in the areas to be dismantled was divided by ten. A decontamination factor of about 300 was obtained, and should allow austenitic steel parts to be melted in special furnaces for unrestricted release. (author)

  5. Data analysis on work activities in dismantling of Japan Power Demonstration Reactor (JPDR). Contract research

    International Nuclear Information System (INIS)

    Shiraishi, Kunio; Sukegawa, Takenori; Yanagihara, Satoshi

    1998-03-01

    The safe dismantling of a retired nuclear power plant was demonstrated by completion of dismantling activities for the Japan Power Demonstration Reactor (JPDR), March, 1996, which had been conducted since 1986. This project was a flag ship project for dismantling of nuclear power plants in Japan, aiming at demonstrating an applicability of developed dismantling techniques in actual dismantling work, developing database on work activities as well as dismantling of components and structures. Various data on dismantling activities were therefore systematically collected and these were accumulated on computer files to build the decommissioning database; dismantling activities were characterized by analyzing the data. The data analysis resulted in producing general forms such as unit activity factors, for example, manpower need per unit weight of component to be dismantled, and simple arithmetic forms for forecasting of project management data to be applied to planning another dismantling project through the evaluation for general use of the analyzed data. The results of data analysis could be usefully applied to planning of future decommissioning of commercial nuclear power plants in Japan. This report describes the data collection and analysis on the JPDR dismantling activities. (author)

  6. Experience of partial dismantling and large component removal of light water reactors

    International Nuclear Information System (INIS)

    Dubourg, M.

    1987-01-01

    Not any of the French PWR reactors need to be decommissioned before the next decade or early 2000. However, feasibility studies of decommissioning have been undertaken and several dismantling scenarios have been considered including the dismantling of four PWR units and the on-site entombment of the active components into a reactor building for interim disposal. In addition to theoretical evaluation of radwaste volume and activity, several operations of partial dismantling of active components and decontamination activities have been conducted in view of dismantling for both PWR and BWR units. By analyzing the concept of both 900 and 1300 MWe PWR's, it appears that the design improvements taken into account for reducing occupational dose exposure of maintenance personnel and the development of automated tools for performing maintenance and repairs of major components, contribute to facilitate future dismantling and decommissioning operations

  7. Progress in the development of tooling and dismantling methodologies for the Windscale advanced gas cooled reactor (WAGR)

    International Nuclear Information System (INIS)

    Cross, M.T.; Wareing, M.I.; Dixon, C.

    1998-01-01

    Decommissioning of the Windscale Advanced Gas-Cooled Reactor (WAGR) is a major UK reactor decommissioning project co-funded by the UK Government, the European Commission and Magnox Electric. WAGR was a CO 2 cooled, graphite moderated reactor which served as a test bed for the development of Advanced Gas-Cooled Reactor technology in the UK. It operated from 1963 until shutdown in 1981. AEA Technology plc are currently the Managing Agents on behalf of UKAEA for the WAGR decommissioning project and are responsible for the co-ordination of the project up to the point when the contents of the reactor core and associated radioactive materials are removed and either disposed of or packaged for disposal at some time in the future. Decommissioning has progressed to the point where the reactor has been dismantled down to the level of the hot gas collection manifold with the removal of the top biological shield, the refuelling standpipes and the top section of the reactor pressure vessel. The 4 heat exchangers have also been removed and committed to shallow land burial. This paper describes the work carried out by AEA Technology under separate contracts of UKAEA in developing some of the equipment and deployment methods for the next phase of active operations required in preparation for the dismantling of the core structure. Most recent work has concentrated on the development of specialist tooling for removal of items of operational waste stored within the reactor core, equipment for cutting and removal of the highly radioactive stainless steel 'loop' pressure tubes, diamond wire cutting equipment for sectioning large diameter pipework, and equipment for dismantling the reactor neutron shield. The paper emphasises the process of adaptation and extension of existing technologies for cost-effective application in the decommissioning environment, the need for adequate forward planning of decommissioning methodologies together with large-scale 'mock-up' testing of equipment to

  8. Deliberated opinion of the Environment Authority concerning the dismantling of the Ulysse reactor (CEA Saclay)

    International Nuclear Information System (INIS)

    2010-01-01

    As the Ulysse reactor of Saclay is about to be dismantled, this report discusses the content of a request made by the CEA to definitely stop and dismantle this reactor. After having recalled the origin of this dismantling project and its regulatory framework, it describes the actual works which are planned, and outlines the main challenges to be faced. It discusses the content of the environmental report or impact study, notably the analysis of the initial condition, the analysis of direct and indirect effects of the project on the environment and health, and of the envisaged measures (waste management, liquid and gaseous effluent management, risk management, radiological exposure of workers in charge of this dismantling, and transports). Recommendations are formulated for a rewriting of the provided documents

  9. Clearance of radioactive materials during reactor dismantling. Permanent enclosure instead of demolition and renaturation?

    International Nuclear Information System (INIS)

    2016-01-01

    During reactor dismantling besides high-level radioactive wastes a large amount of low-level contaminated steel and concrete has to be disposed. In case that radioactivity falls below defined dose limits (10 micro Sv/person and year) these materials may be disposed in domestic waste landfill or in municipal incineration facilities. The issue is discussed in detail including the fact that many power plants are dismantled at the same time so that the contaminated materials might accumulate. Another issue is the occupational safety of contract workers during dismantling. The permanent enclosure could avoid this environmental contamination of decommissioned power plants might also be less expensive.

  10. Remote techniques for the underwater dismantling of reactor internals at the nuclear power plant Gundremmingen unit A

    International Nuclear Information System (INIS)

    Eickelpasch, N.; Steiner, H.; Priesmeyer, U.

    1997-01-01

    Unit A of the nuclear power plant in Gundremmingen (KRB A) is a boiling water reactor with an electrical power of 250 MWe. It was shut down in 1977 after eleven years of operation. The actual decommissioning started in 1983. Since then more than 5200 tons of contaminated components have been dismantled. Special cutting and handling tools were tested, developed and optimized for the purpose of working in radiation fields and under water. Due to the special design of KRB A, which uses a dual-cycle system for additional steam generation, the experience gained is transferable to pressurized water reactors. (Author)

  11. Dismantling of JPDR reactor internals by underwater plasma arc cutting technique using robotic manipulator

    International Nuclear Information System (INIS)

    Yokota, M.

    1988-01-01

    The actual dismantling of JPDR started on December 4, 1986. As of now, equipment that surrounds the reactor has mostly been removed to provide working space in reactor containment prior to the dismantling of reactor internals. Some reactor internals have been successfully dismantled using the underwater arc cutting system with a robotic manipulator during the period of January to March 1988. The cutting system is composed of an underwater plasma arc cutting device and a robotic manipulator. The cut off reactor internals were core spray block, feedwater sparger and stabilizers for fuel upper grid tube. The plasma arc cutting device was developed to dismantle the reactor internals underwater. It mainly consists of a plasma torch, power and gas supply systems for the torch, and by-product treatment systems. It has the cutting ability of 130 mm thickness stainless steel underwater. The robotic manipulator has seven degrees of freedom of movement, enabling it to move in almost the same way as the arm of a human being. The arm of the robot is mounted on a supporting device which is suspended by three chains from the support structure set on a service floor. A plasma torch is griped by the robotic hand; its position to the structure to be cut is controlled from a remote control room, about 100 meters outside the reactor containment

  12. How to replace a reactor pressure vessel

    International Nuclear Information System (INIS)

    Huber, R.

    1996-01-01

    A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

  13. General plan for the partial dismantling of the IRT-Sofia research reactor

    Directory of Open Access Journals (Sweden)

    Apostolov Tihomir G.

    2006-01-01

    Full Text Available After the decision of the Bulgarian Government to reconstruct it, the strategy concerning the IRT-Sofia Research Reactor is to partially dismantle the old systems and equipment. The removal of the reactor core and replacement of old equipment will not pose any significant problems. For a more efficient use of existing resources, there is a need for an engineering project which has been already prepared under the title "General Plan for the Partial Dismantling of Equipment at the IRT-Sofia as a Part of the Reconstruction into a Low Power RR".

  14. The Japan Power Demonstration Reactor (JPDR) dismantling activities. Dismantling of the reactor enclosure and the auxiliary buildings

    International Nuclear Information System (INIS)

    Seiki, Yoshihiro; Kubo, Takashi.

    1996-01-01

    As the final stage of the JPDR decommissioning program, after the major components were removed from each building of JPDR, the dismantling activities proceeded to the decontamination of contaminated concrete surface and the final radiation survey of buildings. These activities were conducted to verify the developed techniques and the detailed procedures for decontamination, and to allow unrestricted use of the JPDR buildings. Following the decontamination of buildings, the dismantling of each building was started. Before dismantling the buildings, the radiation control designations were changed. The buildings that contaminated embedded pipes were changed from first-class radiation controlled areas to second-class radiation controlled areas. On the other hand, the buildings that had no contaminated pipes were changed to uncontrolled areas. A first-class radiation controlled area allows the use of unsealed sources ; thus, radioactive contamination may exist. A second-class radiation controlled area is one where only sealed sources are allowed. Significant quantities of data and experience were obtained during these activities. The practical procedures for decontamination, the final survey of radioactivity, and the dismantling work of buildings were described in this report. (author)

  15. Radioactivity, radiation protection and monitoring during dismantling of light-water reactors

    International Nuclear Information System (INIS)

    Hummel, L.; Zech, J.B.

    2005-01-01

    Based on the radioactivity inventory in the systems and components of light-water reactors observed during operation, the impact of actions during plant emptying after the conclusion of power operation and possible subsequent long-term safe enclosure concerning the composition of the nuclide inventory of the plant to be dismantled will be described. Derived from this will be the effects on radioactivity monitoring in the plant, physical radiation protection monitoring, and the measured characterization of the residual materials resulting from the dismantling. The impact of long-term interim storage will also be addressed in the discussion. The talk should provide an overview of the interrelationships between source terms, decay times and the radioactivity monitoring requirements of the various dismantling concepts for commercial light-water reactors. (orig.)

  16. Stage 2: dismantling of reactor case of the experimental F.B.R. Rapsodie

    International Nuclear Information System (INIS)

    Roger, J.

    1994-01-01

    This document defines the main objectives of stage 2 dismantling of the Rapsodie experimental fast neutron reactor and specifies its time schedule. The work already in progress consists in containing the reactor vessel and its internal equipment, as well as the neutron protection concrete, inside the two leak-tight barriers, and in dismantling all the systems and equipment systems contaminated by sodium. This work, which includes the destruction of 37 metric tons of contaminated sodium from the primary system, was begun in 1987 and will be completed in 1994. The duration of the waiting period for complete dismantling (stage 3) has not been defined. However, the containment and monitoring means implemented should allow a safe waiting period of several decades. (author). 4 figs

  17. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  18. Cold trap dismantling and sodium removal at a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Graf, Anja; Petrick, Holger; Stutz, Uwe [WAK GmbH, Eggenstein-Leopoldshafen (Germany). Hauptabt. Dekontaminationsbetriebe Rueckbau Kompakte Natriumgekuehlte Kernreaktoranlage (KNK); Hosking, Paul [Nuclear Decommissioning Services Limited (NDSL), Sutherland, Dornoch (United Kingdom)

    2013-11-15

    The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, 7 cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed on-site by cutting them up into small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. (orig.)

  19. The management routes for materials produced by the dismantling of the BR3-PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Demeulemeester, Y.; Ponnet, M.; Emond, M.; Emond, O.; Dadoumont, J.; Massaut, V.

    2000-01-01

    The dismantling of the BR3 reactor produces quite large masses of contaminated materials, mainly metals or concrete. The main management routes are: conditioning of the radioactive wastes and disposal, recycling of radioactive materials in the nuclear sector and the recycling of free released materials in the industrial sector or their evacuation as industrial waste. The conditioning of the radioactive wastes is essentially performed in the installations of Belgoprocess and must follow the specifications imposed by the national radwaste management agency ONDRAF/NIRAS. The conditioning of the pieces produced during the cutting of the reactor pressure vessel is given as example. The recycling of radioactive materials in the nuclear sector is possible for metals and for concrete. For metals, SCK.CEN has an agreement with a nuclear foundry which reuses these materials for the fabrication of shieldings. For concrete, an R and D programme is going on with the objective to demonstrate the possible reuse of baryte concrete as raw materials for the production of mortar used in the conditioning of radioactive wastes. The free release of radioactive materials and their reuse or evacuation as radioactive wastes requires the strict respect of procedures and the use of low level measurement techniques. Various decontamination techniques are used at SCK.CEN to reach this objective. For the metals, we use mainly simple washing, abrasive decontamination and hard chemical decontamination. For concrete, we use mainly scabbling or shaving techniques. (authors)

  20. Remote dismantlement tasks for the CP5 reactor: Implementation, operations, and lessons learned

    International Nuclear Information System (INIS)

    Noakes, M.W.

    1998-01-01

    This paper presents a developer's perspective on lessons learned from one example of the integration of new prototype technology into a traditional operations environment. The dual arm work module was developed by the Robotics Technology Development Program as a research and development activity to examine manipulator controller modes and deployment options. It was later reconfigured for the dismantlement of the Argonne National Laboratory Chicago Pile number-sign 5 reactor vessel as the crane-deployed dual arm work platform. Development staff worked along side operations staff during a significant part of the deployment to provide training, maintenance, and tooling support. Operations staff completed all actual remote dismantlement tasks. At the end of available development support funding, the Dual Arm Work Platform was turned over to the operations staff, who is still using it to complete their dismantlement tasks

  1. Remote dismantling of the French Brennilis nuclear power plant

    International Nuclear Information System (INIS)

    Studenski, Joerg

    2009-01-01

    The paper deals with the remote dismantling of the decommissioned EL4 prototype power plant Brennilis in France. The block contains the reactor pressure vessel including internals and biological shield, the piping and the control systems. The authors describe the general operation principle of the reactor to illustrate the peculiarities of the dismantling concept and the concept-related challenges. Detailed information is given concerning the following issues: creation of an access to the reactor block, the used remote technology, dismantling of the coolant piping and the axial shield, dismantling of the reactor pressure vessel and the lateral shield. Special attention is given on the minimization of the produced radioactive waste.

  2. Remote dismantling of the French Brennilis nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Studenski, Joerg [NUKEM Technologies GmbH (Germany)

    2009-07-01

    The paper deals with the remote dismantling of the decommissioned EL4 prototype power plant Brennilis in France. The block contains the reactor pressure vessel including internals and biological shield, the piping and the control systems. The authors describe the general operation principle of the reactor to illustrate the peculiarities of the dismantling concept and the concept-related challenges. Detailed information is given concerning the following issues: creation of an access to the reactor block, the used remote technology, dismantling of the coolant piping and the axial shield, dismantling of the reactor pressure vessel and the lateral shield. Special attention is given on the minimization of the produced radioactive waste.

  3. Experience of the remote dismantling of the Windscale advanced gas-cooled reactor and Windscale pile chimneys

    International Nuclear Information System (INIS)

    Wright, E.M.

    1993-01-01

    This paper gives brief descriptions of some of the remote dismantling work and equipment used on two large decommissioning projects: the BNFL Windscale Pile Chimneys Project (remote handling machine, waste packaging machine, remotely controlled excavator, remotely controlled demolition machine) and the AEA Windscale Advanced Gas-cooled Reactor Project (remote dismantling machine, operational waste, bulk removal techniques, semi-remote cutting operations)

  4. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

    Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

  5. Project WAGR: the UK demonstration project for power reactor decommissioning - a review of the tools used to dismantle the reactor core

    International Nuclear Information System (INIS)

    Benest, T.G.

    2008-01-01

    The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940. UKAEA mission is to restore the environment of its sites in a safe and secure manner. This restoration includes the decommissioning of a number of redundant research and power reactors. The Windscale Advanced Gas-cooled Reactor (WAGR) was the UK prototype Advanced gas cooled reactor and became the forerunner of a family of 14 reactors built to generate cheaper and more efficient electricity in the UK. WAGR was constructed between 1957 and 1961 and was a carbon dioxide cooled, graphite moderated reactor using uranium oxide fuel in stainless steel cans. The reactor consisted of a graphite moderator housed in a cylindrical reactor vessel with hemispherical ends. The reactor and associated heat exchangers were enclosed in the iconic spherical containment building regularly used by the media in the UK as an illustration of the nuclear industry. The reactor first produced power in August 1962 and achieved full design output in 1963. It operated at an electrical output of 33 MW (E) for 18 years (average load factor of 75%). In 1981 the reactor was shut down after satisfactory completion of all the research and development objectives. In anticipation of the UK likely nuclear decommissioning needs the UKAEA decided to decommission WAGR to the International Atomic Energy Agency (IAEA) stage 3 (restoration of the area occupied by the facility to a condition of unrestricted re-usability) as the national demonstration exercise for power reactor decommissioning. Since 1998 the UKAEA and its contractors have been undertaking the dismantling of the reactor core components and pressure vessel in a series of 10 campaigns. These contain neutron activated components expected to produce dose rates well in excess of 1 Sv/hr. To carry out the work UKAEA installed an 8M remote dismantling machine (RDM) a waste recovery and transport system and a shielded waste

  6. The Molten Salt Reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.; Dodds, H.L.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined. It is concluded that MSRs are very suitable for beneficial utilization of the dismantled fuel. The MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus MSRs are flexible while maintaining their economy. MSRs further require a minimum of special fuel preparation and can tolerate denaturing and dilution of the fuel. Fuel shipments can be arbitrarily small, all of which supports nonproliferation and averts diversion. MSRs have inherent safety features which make them acceptable and attractive. They can burn a fuel type completely and convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems for deployment of nuclear power. 19 refs

  7. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2003-07-01

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

  8. State of dismantling of reactor facilities (JPDR) in Tokai Research Establishment, Japan Atomic Energy Research Institute (fiscal year 1986)

    International Nuclear Information System (INIS)

    1987-01-01

    As to this dismantling work, the contents of the notice on the dismantling and the policy of administration offices to deal with it were reported to the Nuclear Safety Commission beforehand, and after the approval was obtained, it has been executed. Based on the talk of the chairman of Nuclear Safety Commission on January 6, 1983, the state of dismantling is reported. In fiscal year 1986, the second stage dismantling was begun in December, and it was confirmed that the works were carried out safety. In order to ensure the place required for dismantling and removing in-core structures and the pressure vessel, the pressure vessel upper cover and its heat insulator were removed from the third story of the containment vessel. A part of the facilities installed on the second and third stories of the containment vessel was also dismantled and removed. In order to utilize as the place for the temporary preservation and decontamination of dismantled things, a part of the facilities installed in the dump condenser building was dismantled and removed. The control building was reconstructed for smoothly carrying out the entrance control of workers. 36 spent fuel assemblies were carried away for reprocessing. The exposure dose of workers was below the detectable limit of film badges. The dismantled wastes were about 171 t of metals and 34 t of concrete. (Kako, I.)

  9. Reactor pressure vessel support

    International Nuclear Information System (INIS)

    Butti, J.P.

    1977-01-01

    A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

  10. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  11. Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

    International Nuclear Information System (INIS)

    ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

    2017-01-01

    The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

  12. Reactor pressure vessel design

    International Nuclear Information System (INIS)

    Foehl, J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

  13. Adaptation of high pressure water jets with abrasives for nuclear installations dismantling

    International Nuclear Information System (INIS)

    Rouviere, R.; Pinault, M.; Gasc, B.; Guiadeur, R.; Pilot, M.

    1989-01-01

    This report presents the work realized for adjust the cutting technology with high pressure water jet with abrasives for nuclear installation dismantling. It has necessited the conception and the adjustement of a remote tool and the realization of cutting tests with waste produce analysis. This technic can be ameliorated with better viewing systems and better fog suction systems

  14. Latest feedback from a major reactor vessel dismantling project

    International Nuclear Information System (INIS)

    Boucau, J.; Segerud, P.; Sanchez, M.; Garcia, R.

    2015-01-01

    Westinghouse performed two large segmentation projects in 2010-2013 and then 2013-2015 at the Jose Cabrera nuclear power plant in Spain. The power plant is located in Almonacid de Zorita, 43 miles east of Madrid, Spain and was in operation between 1968 and 2006. This paper will describe the sequential steps required to prepare, segment, separate, and package the individual component segments using under water mechanical techniques. The paper will also include experiences and lessons learned that Westinghouse has collected from the activities performed during the reactor vessel and vessel internals segmentation projects. (authors)

  15. Radiation exposure of the personnel during dismantling and cutting of the primary system of the Karlsruhe Multi-purpose Research Reactor (MZFR)

    International Nuclear Information System (INIS)

    Hesse, H.; Demant, W.; Reichert, A.; Willmann, F.

    2000-10-01

    The heavy water (D 2 O) cooled and moderated pressurized water reactor MZFR with a thermal power of 200 MW will be dismantled step-by-step within the framework of sectional decommissioning licenses. The past decommissioning step (6 th sectional license) in general covered the removal of the primary systems and of all reactor support systems inside the reactor building. The measures for radiation protection during dismantling and handling of the large components of the primary system, such as the fuel element loading machine, fuel element transfer system, steam generator and pressurizer shall be pointed out. The measures taken for the reduction of the dose rate during dismantling and cutting of the components for the purpose of conditioning or unrestricted reuse at the central decontamination department (HDB) shall be described. Chemical decontamination of the primary circuit and its components, which had to be executed in order to reduce the dose rates for subsequent manual dismantling, shall be presented. The efforts undertaken for the protection of individuals in view of the difficult radiological boundary conditions (high concentrations of tritium in all systems as well as very high alpha contamination) will be explained. Moreover, dose-minimizing measures during cutting of the primary circuit and its components at HDB shall be described by the example of the cutting of a steam generator. It shall be demonstrated that cutting and dismantling of highly contaminated and activated parts with high dose rates can be executed safely in terms of both the radiation exposure of the personnel and the technical, financial and time expenditure. (orig.)

  16. Reactor pressure tank

    International Nuclear Information System (INIS)

    Dorner, H.; Scholz, M.; Jungmann, A.

    1975-01-01

    In a reactor pressure tank for a nuclear reactor, self-locking hooks engage a steel ring disposed over the removable cover of the steel vessel. The hooks exert force upon the cover to maintain the cover in a closed position during operation of the reactor pressure tank. The force upon the removal cover is partly the result of the increasing temperature and thermal expansion of the steel vessel during operation. The steel vessel is surrounded by a reinforced-concrete tank. (U.S.)

  17. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    Hong, Jun Hwa; Chi, S. H.; Lee, B. S.

    2002-04-01

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  18. Decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East. Project final report

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Clark, F.R.

    1997-10-01

    The decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East (ANL-E) was completed in October 1997. Descriptions and evaluations of the activities performed and analyses of the results obtained during the JANUS D and D Project are provided in this Final Report. The following information is included: objective of the JANUS D and D Project; history of the JANUS Reactor facility; description of the ANL-E site and the JANUS Reactor facility; overview of the D and D activities performed; description of the project planning and engineering; description of the D and D operations; summary of the final status of the JANUS Reactor facility based upon the final survey results; description of the health and safety aspects of the project, including personnel exposure and OSHA reporting; summary of the waste minimization techniques utilized and total waste generated by the project; and summary of the final cost and schedule for the JANUS D and D Project

  19. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

    Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

  20. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  1. Decontamination and radioactivity measurement on building surfaces related to dismantling of Japan power demonstration reactor (JPDR)

    International Nuclear Information System (INIS)

    Hatakeyama, Mutsuo; Tachibana, Mitsuo; Yanagihara, Satoshi

    1997-12-01

    In the final stage of dismantling activities for decommissioning a nuclear power plant, building structures have to be demolished to release the site for unrestricted use. Since building structures are generally made from massive reinforced concrete materials, it is not a rational way to treat all concrete materials arising from its demolition as radioactive waste. Segregation of radioactive parts from building structures is therefore indispensable. The rational procedures were studied for demolition of building structures by treating arising waste as non-radioactive materials, based on the concept established by Nuclear Safety Commission, then these were implemented in the following way by the JPDR dismantling demonstration project. Areas of the JPDR facilities are categorized into two groups : possibly contaminated areas, and possibly non-contaminated areas, based on the document of the reactor operation. Radioactivity on the building surfaces was then measured to confirm that the qualitative categorization is reasonable. After that, building surfaces were decontaminated in such a way that the contaminated layers were removed with enough margin to separate radioactive parts from non-radioactive building structures. Thought it might be possible to demolish the building structures by treating arising waste as non-radioactive materials, confirmation survey for radioactivity was conducted to show that there is no artificial radioactive nuclides produced by operation in the facility. This report describes the procedures studied on measurement of radioactivity and decontamination, and the results of its implementation in the JPDR dismantling demonstration project. (author)

  2. Pressure tube reactor

    International Nuclear Information System (INIS)

    Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

    1993-01-01

    In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

  3. EPR (European Pressurized Reactor)

    International Nuclear Information System (INIS)

    2015-01-01

    This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

  4. Pressure tube reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Fujino, Michihira.

    1980-01-01

    Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

  5. Robotic dismantlement systems at the CP-5 reactor D and D project

    International Nuclear Information System (INIS)

    Seifert, L. S.

    1998-01-01

    The Chicago Pile 5 (CP-5) Research Reactor Facility is currently undergoing decontamination and decommissioning (D and D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principle nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water cooled and moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D and D project includes the disassembly, segmentation and removal of all the radioactive components, equipment and structures associated with the CP-5 facility. The Department of Energy's Robotics Technology Development Program and the Federal Energy Technology Center, Morgantown Office provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor assembly for tasks requiring remote dismantlement as part of the EM-50 Large-Scale Demonstration Program (LSDP). The teleoperated systems provided were the Dual Arm Work Platform (DAWP), the Rosie Mobile Teleoperated Robot Work System (ROSIE), and a remotely-operated crane control system with installed swing-reduction control system. Another remotely operated apparatus, a Brokk BM250, was loaned to ANL by the Princeton Plasma Physics Laboratory (PPPL). This machine is not teleoperated and was not part of the LSDP, but deserves some mention in this discussion. The DAWP is a robotic dismantlement system that includes a pair of Schilling Robotic Systems Titan III hydraulic manipulator arms mounted to a specially designed support platform: a hydraulic power unit (HPU) and a remote operator console. The DAWP is designed to be crane-suspended for remote positioning. ROSIE, developed by RedZone Robotics, Inc. is a mobile, electro-hydraulic, omnidirectional platform with a heavy-duty telescoping boom mounted to the platform's deck. The work system includes the mobile platform (locomotor), a power distribution unit (PDU) and a remote operator console. ROSIE moves about the reactor building

  6. Optimization study and preliminary design for Latina NPP early core retrieval and reactor dismantling

    International Nuclear Information System (INIS)

    Macci, E.; Zirpolo, S.; Imparato, A.; Cacace, A.; Parry, D.; Walkden, P.

    2002-01-01

    In June 2000, an agreement was established between Sogin and BNFL to enable the two companies to co-operate, using their specific experiences in the decommissioning field, for the benefit of projects in Italy, the United Kingdom and for third markets. A decommissioning strategy for the Latina NPP was initially developed in a Phase 1 Study which produced a conceptual design for the decommissioning of the reactor. This study was completed in June 2000. Since then, a second study has been completed, which has further developed the strategy and produced preliminary designs for the early dismantling of the core and reactor building at Latina. The engineering and safety data were produced in order to support Sogin in the preparation of a safety case for plant decommissioning. This safety case was submitted to the Italian Regulator, ANPA, in February 2002. (author)

  7. The molten salt reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined and is found very suitable for the beneficial use of this fuel. MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus, MSRs are flexible while maintaining their economy. Furthermore, MSRs require only a minimum of special fuel preparation. They can tolerate denaturing and dilution of their fuel. The size of fuel shipments can be determined to optimize safety and security-all of which supports nonproliferation and resists diversion. In addition, MSRs have inherent safety features that make them acceptable and attractive. They can burn fissile material completely or can convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems in the deployment of nuclear power

  8. Feedback from dismantling operations (level 2) on EDF's first generation reactors

    International Nuclear Information System (INIS)

    West, J P.; Dionisio-Gomes, A.; Kus, J P.; Mervaux, P.; Bernet, P.; Dalmas, R.

    2003-01-01

    EDF's policy as regards the dismantling of the reactors that have ceased commercial operation, namely the eight power plants of the first generation and the Creys-Malville power plant, is explained. Generally speaking, prior to the year 2001, EDF had opted for the de-construction of these power plants to comply with a 'long wait' scenario, which consisted of waiting for a period of 5 to 10 years to achieve IAEA level 2 (partial release of the site), then postponing the total de-construction of the facilities for 25 to 50 years. Today, EDF has decided to undertake the total de-construction of these reactors, which have ceased commercial operation, over a period of 25 years. The purpose of this document is to present: - The reactors concerned, their background and their 'regulatory' situation, - The main operations performed and/or currently in progress, - The main elements of feedback from such operations, shedding light on the approach adopted in 2001. The installations concerned by the de-construction programme are as follows: - The 8 power plants of the first generation, which were built during the fifties and sixties and ceased commercial operation between 1973 and 1994, namely: Brennilis (industrial prototype using heavy water technology, jointly operated by EDF and CEA), the 6 power units of the NUGG type (natural uranium gas graphite) at Chinon, Saint-Laurent des Eaux and Bugey and the PWR reactor at Chooz A, - The storage silos at Saint-Laurent, where the sleeves for the fuel assemblies of reactors SLA1 and SLA2 are stored, corresponding to approximately 2000 tonnes of graphite, - The Creys-Malville reactor, FBR (fast breeder reactor) shut down in accordance with a government decision, which is currently undergoing decommissioning. At the current stage, our feedback from the dismantling operations carried out on nuclear facilities is based on (i) the work carried out or in progress that will make it possible to achieve the equivalent of IAEA level 2 in the

  9. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  10. Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

    Science.gov (United States)

    Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

    2018-05-01

    The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

  11. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-01-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  12. Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

    1998-07-01

    SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

  13. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  14. Characterization of radioactive graphite and concrete of the reactor ULYSSE/INSTN at CEA/Saclay to be dismantled

    International Nuclear Information System (INIS)

    Van Lauwe, Aymeric; Ridikas, Danas; Damoy, Francois; Blideanu, Valentin; Fajardo, Christophe; Aubert, Marie-Cecile; Foulon, Francois

    2006-01-01

    Decommissioning and dismantling of nuclear installations after their service life are connected with the necessity of the disassembling, handling and disposing of a large amount of radioactive material. In order to optimize the disassembling operations, to reduce the undesirable volume to the minimum and to successfully plan the dismantling and disposal of radioactive materials to storage facilities, the radiological characterisation of the material present in the reactor and around its environment should be accurately evaluated. The present work has been done in the framework of the decommissioning and dismantling of the experimental reactor ULYSSE that is presently operating in INSTN/Saclay and will be closed in the middle of 2006. A methodology, already successfully used for another research reactor, is proposed for determining accurately the long-term induced activity of the materials present in the active reactor core and its surroundings. The comparison of theoretical predictions, based on Monte Carlo technique, with experimental values validated the approach and the methodology used in the present study. The goal is to plan efficiently the disassembling and dismantling of the system and to optimise the mass flow going to different waste repositories. We show that this approach might reduce substantially the total cost of decommissioning. (authors)

  15. Declassification of radioactive water from a pool type reactor after nuclear facility dismantling

    Science.gov (United States)

    Arnal, J. M.; Sancho, M.; García-Fayos, B.; Verdú, G.; Serrano, C.; Ruiz-Martínez, J. T.

    2017-09-01

    This work is aimed to the treatment of the radioactive water from a dismantled nuclear facility with an experimental pool type reactor. The main objective of the treatment is to declassify the maximum volume of water and thus decrease the volume of radioactive liquid waste to be managed. In a preliminary stage, simulation of treatment by the combination of reverse osmosis (RO) and evaporation have been performed. Predicted results showed that the combination of membrane and evaporation technologies would result in a volume reduction factor higher than 600. The estimated time to complete the treatment was around 650 h (25-30 days). For different economical and organizational reasons which are explained in this paper, the final treatment of the real waste had to be reduced and only evaporation was applied. The volume reduction factor achieved in the real treatment was around 170, and the time spent for treatment was 194 days.

  16. Pressure tube reactors

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1981-01-01

    Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

  17. Activation calculations for dismantling - The feedback of a 7 years experience in activation calculations for graphite gas cooled reactors in France

    International Nuclear Information System (INIS)

    Eid, M.; Nimal, J.C.; Gerat, L.M.

    1994-01-01

    This is a revision of the past seven years experience in activation calculations for dismantling. It aims at evaluating the experience and at making better understanding to help in decision making during the following phases. Five gas cooled reactors are shutdown and are waiting for the EDF (Electricite De France) dismantling decision. The sixth (BUGEY1) will be shutdown by 1994 and will be waiting a dismantling decision as well. (authors). 3 figs., 3 tabs

  18. CP-5 reactor remote dismantlement activities: Lessons learned in the integration of new technology in an operations environment

    International Nuclear Information System (INIS)

    Noakes, M.W.

    1998-01-01

    This paper presents the developer's perspective on lessons learned from one example of the integration of new prototype technology into a traditional operations environment. The dual arm work module was developed by the Robotics Technology Development Program as a research and development activity to examine manipulator controller modes and deployment options. It was later reconfigured for the dismantlement of the Argonne National Laboratory Chicago Pile No. 5 reactor vessel as the crane-deployed dual arm work platform. Development staff worked along side operations staff during a significant part of the deployment to provide training, maintenance, and tooling support. Operations staff completed all actual remote dismantlement tasks. At the end of available development support funding, the Dual Arm Work Platform was turned over to the operations staff, who are still using it to complete their dismantlement tasks

  19. Decommissioning, Dismantling and Disarming: a Unique Information Showroom Inside the G2 Reactor at Marcoule Centre (France) - 12068

    Energy Technology Data Exchange (ETDEWEB)

    Volant, Emmanuelle [CEA DAM, Bruyeres-le-Chatel (France); Garnier, Cedric [CEA DEN, Marcoule (France)

    2012-07-01

    The paper aims at presenting the new information showroom called 'Escom G2' (for 'Espace Communication') inaugurated by the French Atomic Energy and Alternative Energies Commission (CEA) in spring 2011. This showroom is settled directly inside the main building of the G2 nuclear reactor: a facility formerly dedicated to weapon-grade plutonium production since the late 1950's at the Marcoule nuclear centre, in south of France. After its shutdown, and reprocessing of the last spent fuels, a first dismantling step was successfully completed from 1986 to 1996. Unique in France and in Europe, Escom G2 is focused on France dismantling expertise and its action for disarmament. This showroom comprises of a 300-square meters permanent exhibition, organized around four themes: France strategy for disarmament, decommissioning and dismantling technical aspects, uranium and plutonium production cycles. Each of these topics is illustrated with posters, photos, models and technical pieces from the dismantled plants. It is now used to present France's action in disarmament to highly ranked audiences such as: state representatives, diplomats, journalists... The paper explains the background story of this original project. As a matter of fact, in 1996 France was the first nuclear state to decide to shut down and dismantle its fissile material production facilities for nuclear weapons. First, the paper presents the history of the G2 reactor in the early ages of Marcoule site, its operating highlights as well as its main dismantling operations, are presented. In Marcoule, where the three industrial-scale reactors G1, G2 and G3 used to be operated for plutonium production (to be then reprocessed in the nearby UP1 plant), the initial dismantling phase has now been completed (in 1980's for G1 and in 1996 for G2 and G3). The second phase, aimed at completely dismantling these three reactors, will restart in 2020, and is directly linked to the opening of

  20. Melting of contaminated steel scrap from the dismantling of the CO2 systems of gas cooled, graphite moderated nuclear reactors

    International Nuclear Information System (INIS)

    Feaugas, J.; Jeanjacques, M.; Peulve, J.

    1994-01-01

    G2 and G3 are the natural Uranium cooled reactors Graphite/Gas. The two reactors were designed for both plutonium and electricity production (45 MWe). The dismantling of the reactors at stage 2 has produced more than 4 000 tonnes of contaminated scrap. Because of their large mass and low residual contamination level, the French Atomic Energy Commission (CEA) considered various possibilities for the processing of these metallic products in order to reduce the volume of waste going to be stored. After different studies and tests of several processes and the evaluation of their results, the choice to melt the dismantled pipeworks was taken. It was decided to build the Nuclear Steel Melting Facility known as INFANTE, in cooperation with a steelmaker (AHL). The realization time schedule for the INFANTE lasted 20 months. It included studies, construction and the licensing procedure. (authors). 2 tabs., 3 figs

  1. Aspects of reactor dismantling planning following the safe entombment in the NPP Lingen (KWL); Aspekte der Abbauplanung nach dem Sicheren Einschluss im Kernkraftwerk Lingen (KWL)

    Energy Technology Data Exchange (ETDEWEB)

    Priesmeyer, U.; Rojahn, T.; Fries, B. [Kernkraftwerk Lingen GmbH (Germany)

    2009-07-01

    The NPP Lingen (KWL) was shut-down in 1977. Due to the fact that no final repository was available the safe entombment for 25 years was chosen following the decommissioning. The conventional plant components were dismantled and removed from the plant site. The licensing procedure for reactor dismantling with final disposal in Schacht Konrad has been started. The beginning of dismantling operation is scheduled for 2013. The authors describe the preparatory work, the boundary conditions for the dismantling, radiation protection considerations with respect to manual demolition work after the rather long decay time.

  2. Reactor pressure vessel embrittlement

    International Nuclear Information System (INIS)

    1992-07-01

    Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

  3. Economical dismantling of nuclear power stations

    International Nuclear Information System (INIS)

    Mallok, J.; Andermann, H.

    1999-01-01

    The dismantling of nuclear power stations requires a high degree of security and economic efficiency due to the strong contamination of components and the close spatial conditions. In order to protect involved staff from radiation, modern remote-controlled technology is applied in sectors with heavy radioactive contamination such as reactor pressure vessels. The article shows, that the dismantling of reactor pressure vessels using a remote-controlled milling machine developed by the Siemens subsidiary Mechanik Center Erlangen GmbH, can be done in a secure and efficient way. (orig.) [de

  4. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1985-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

  5. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1980-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

  6. The dismantling of the damaged Chernobyl reactor No 4 and the implications upon the design of a second shelter

    International Nuclear Information System (INIS)

    Mao, J.L. Le; White, S.J.

    1995-01-01

    During 1994/95 a study was undertaken, by Alliance, to assess the feasibility of constructing a new containment/shelter over the damaged Chernobyl reactor and its sarcophagus. This paper reviews the implications placed upon the design due to the ultimate dismantling requirements. The paper highlights some of the generic requirements and the impact upon the shelter design solution. Alliance is a consortium of leading European engineering companies comprising Campenon Bernard, AEA Technology, Walter-Bau, SGN, Taylor Woodrow and Bouygues. (Author)

  7. Dismantling of the research reactor RTS-1 Galileo Galilei in Pisa (Italy)

    International Nuclear Information System (INIS)

    Ruiz Martinez, J. t.; Farella, G.; Cimini, E.; Russo, M.

    2014-01-01

    This paper is about the most relevant aspects of the first phase of the dismantling, removal of the water in the pool, prior treatment through evaporation, the dismantling of all the submerged activated elements and other activated or contaminated elements that have been part of the nuclear facility. (Author)

  8. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  9. Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

    International Nuclear Information System (INIS)

    Perry, E.F.

    1995-01-01

    The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

  10. Remote control for the underwater dismantling of reactor internals at the nuclear power plant Gundremmingen unit A

    International Nuclear Information System (INIS)

    Eickelpasch, N.; Steiner, H.; Priesmeyer, U.

    1996-01-01

    The unit A of the nuclear power plant in Gundremmingen (KRB A) is a boiling water reactor with an electrical power of 250 MW e . It was shut down in 1977 after 11 years of operation. The actual decommissioning started in 1983. Meanwhile more than 5200 tons of contaminated components have been dismantled. Special cutting and handling tools were tested, developed and optimized for the purpose of working in radiation fields and under water. Due to the special design of KRB A, using an dual cycle system for additional steam generation, the experience gained is transferable to pressurised water reactors as well. (Author)

  11. Challenges of dismantling

    International Nuclear Information System (INIS)

    Chevet, P.F.; Schilz, F.; Rondeau, J.M.; Piketty, L.; Dupraz, B.; Conte, D.; Duguey, M.; Louet, C.A.; Dorison, A.; Dutzer, M.; Boucau, J.; Eimer, M.; Boutin, D.; Revilla, J.L.; Golshan, M.; Smith, G.

    2015-01-01

    This document is made up of short articles whose issue is reactor dismantling. The first article presents the French strategy that can be featured by immediate dismantling (the dismantling process is prepared a long time before decommissioning and begins as soon as the reactor is shut down) and massive dismantling (a lot of nuclear facilities will be decommissioned in a near future). The following 4 articles give the viewpoints of ASN (Nuclear Safety Authority), EDF (for its fleet of PWRs), CEA (for its experimental reactors and nuclear facilities) and AREVA (for the EURODIF George Besse plant). Costs and financing are dealt with in an article that says that the cost is greatly dependent on the final state: a complete nuclear-free area or an area whose radioactivity is below safe standards and that law implies to constitute provisions all along the operating life of the facility to cover dismantling costs. Dismantling generates a huge amount of very low-level radioactive wastes particularly metal scraps that might be recycled and get out of nuclear industry, an article details the feasibility of such recycling. Another article shows the impact of massive dismantling on the management of radioactive wastes. In an article Westinghouse presents its experience in the cutting of internal equipment of the reactor core. The last 2 articles presents the dismantling strategies in Spain and in the UK. (A.C.)

  12. Optimized phases for reactor dismantling – an efficient and sustainable concept

    International Nuclear Information System (INIS)

    Krüger, S.; Winter, J.

    2013-01-01

    D&D projects are driven by costs, to implement an optimization process from the very beginning is key. Optimized strategy and sequencing of the dismantling (hot to cold) will provide serious economical savings . Larger dismantling packages will reduce interfaces and ease the coordination efforts on site. Early usage of mobile systems will ease the large-scale release for dismantling Social transition has to be addressed with priority and to be planned at an early phase in the D&D planning Concept, Planning & Project Management will influence the success of the project much more than the used technique

  13. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  14. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  15. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

    Strosnider, J.; Wichman, K.; Elliot, B.

    1994-12-01

    This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

  16. Clearance of radioactive materials during reactor dismantling. Permanent enclosure instead of demolition and renaturation?; Freigabe radioaktiven Materials beim AKW-Abriss. Dauerhafter Einschluss statt Rueckbau?

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-07-01

    During reactor dismantling besides high-level radioactive wastes a large amount of low-level contaminated steel and concrete has to be disposed. In case that radioactivity falls below defined dose limits (10 micro Sv/person and year) these materials may be disposed in domestic waste landfill or in municipal incineration facilities. The issue is discussed in detail including the fact that many power plants are dismantled at the same time so that the contaminated materials might accumulate. Another issue is the occupational safety of contract workers during dismantling. The permanent enclosure could avoid this environmental contamination of decommissioned power plants might also be less expensive.

  17. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

    2017-11-15

    One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

  18. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

  19. Technology, safety, and costs of decommissioning a reference pressurized water reactor power station

    International Nuclear Information System (INIS)

    Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.

    1978-05-01

    Safety and cost information was developed for the conceptual decommissioning of a large [1175 MW(e)] pressurized water reactor (PWR) power station. Two approaches to decommissioning, Immediate Dismantlement and Safe Storage with Deferred Dismantlement, were studied to obtain comparisons between costs, occupational radiation doses, potential radiation dose to the public, and other safety impacts. Immediate Dismantlement was estimated to require about six years to complete, including two years of planning and preparation prior to final reactor shutdown, at a cost of $42 million, and accumulated occupational radiation dose, excluding transport operations, of about 1200 man-rem. Preparations for Safe Storage were estimated to require about three years to complete, including 1 1 / 2 years for planning and preparation prior to final reactor shutdown, at a cost of $13 million and an accumulated occupational radiation dose of about 420 man-rem. The cost of continuing care during the Safe Storage period was estimated to be about $80 thousand annually. Accumulated occupational radiation dose during the Safe Storage period was estimated to range from about 10 man-rem for the first 10 years to about 14 man-rem after 30 years or more. The cost of decommissioning by Safe Storage with Deferred Dismantlement was estimated to be slightly higher than Immediate Dismantlement. Cost reductions resulting from reduced volumes of radioactive material for disposal, due to the decay of the radioactive containments during the deferment period, are offset by the accumulated costs of surveillance and maintenance during the Safe Storage period

  20. Development of plasma arc cutting technique for dismantlement of reactor internals in JPDR decommissioning program

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi; Tanaka, Mitsugu; Ujihara, Norio.

    1988-01-01

    The decommissioning program for JPDR has been conducted by JAERI since 1981 under contact with the Science and Technology Agency of Japan. The development of cutting tools for dismantling the JPDR is one of the important items in the program. An underwater plasma arc cutting technique was selected for dismantling the JPDR core internals. The study was concentrated on improving the cutting ability in water. Various cutting tests were conducted changing the parameters such as arc current, supply gas and cutting speed to evaluate the most effective cutting condition. Through the study, it has been achieved to be able to cut a 130 mm thick stainless steel plate in water. In addition, the amount and the characteristics of by-products were measured during the cutting tests for the safety evaluation of the dismantling activities. Final cutting tests and checkout of whole plasma arc cutting system were conducted using a mockup water pool and test pieces simulating the JPDR core internals. It was proved from the tests that the cutting system developed in the program will be applicable for the JPDR core internals dismantlement. (author)

  1. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

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

  2. Pressure tube reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Kaneto, Kunikazu.

    1979-01-01

    Purpose: To attain uniform fluid poison distribution in a calandria tank by downwardly projecting, at an equal distance to the reactor core, a spacer wall from the periphery of an anti-vibration plate in the vicinity of a heavy water flow passage in the periphery of the anti-vibration plate, thereby decrease the amount of heavy water flowing into the heavy water flow passage. Constitution: A projecting wall concentrical with a calandria tank is suspended vertically from the boundary side at the peripheral portion of an anti-vibration plate to a water heavy flow passage in the periphery of the anti-vibration plate. The projecting wall has such a vertical length as about equal to the width of the heavy water flow passage, prevents heavy water flowing through apertures of a control rod guide tube from entering into the heavy water passage and increases the ratio of heavy water that flows through the heavy water flow passage in the anti-vibration plate. Consequently, if the liquid poison density in heavy water is varied, the ununiform poison density in the calandria tank can be prevented. (Seki, T.)

  3. The Superphenix dismantling

    International Nuclear Information System (INIS)

    Carle, R.

    1999-01-01

    This document presents selected abstracts of Remy Carle's presentation on the dismantling of Superphenix (october 1998). The author wonders about the consequences of such a decision. After a chronological account of this fast reactor project, its cost and the scientific and technical contribution, the dismantling problem is considered. For EDF (Electricite De France) the dismantling dimension is considered at the same time of the design. The main problem is the liquid sodium reprocessing: a technical but also a financing problem. The end of the speech deals with the political aspects of Superphenix and the relations with the public. (A.L.B.)

  4. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

    Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

  5. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

  6. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  7. Pressurized water reactor with a reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1979-01-01

    The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

  8. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

  9. Comparison of thorough decontamination techniques on dismantled pieces of a PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Rahier, A.; Mandoki, R.; Ponnet, M.

    1998-01-01

    The decontamination experience gained during the BR3 dismantling project is developed. This started with the full system decontamination of the primary loop and was followed by R and D on thorough decontamination projects. First, a wet abrasive installation has been installed and is now in operation for the thorough cleaning of metallic pieces of simple geometry. Afterwards, the chemical cerium process has been developed. The results of the regeneration with ozone and with electrochemistry are presented in detail. The ozone regeneration process has been selected for the industrial installation of which the construction is foreseen in 1998. (author)

  10. Sampling of reactor pressure vessel and core internals

    International Nuclear Information System (INIS)

    Oberhaeuser, Ralf

    2012-01-01

    Decommissioning and dismantling of nuclear power plants is a growing business as a huge number of plants built in the 1970's have now reached their lifetime. It is well known that dismantling a nuclear power plant means an extraordinary expense for the owner respectively operator. Beside the dismantling works for itself, the disposal of activated components and other nuclear waste is very expensive. What comes next is the fact that final disposal facilities are not available yet in most countries meaning a need for interim storage on-site in specially built facilities. It can be concluded that a special attention is paid on producing a minimal radioactive waste volume. For this, optimized dismantling and packaging concepts have to be developed. AREVA is proud of versatile experience in successfully dismantling nuclear components like core internals and reactor pressure vessel (RPV). The basis of a well-founded and optimized dismantling and packaging concept must always be the detailed knowledge of the radiological condition of the component to be and in the best case a 3D activation- model. For keeping the necessary sampling effort as small as possible, but simultaneously as efficient as possible, representative sampling positions are defined in advance by theoretical radiological examinations. For this, a detailed 3D-CAD-model of the components to be dismantled has proven very helpful and effective. Under these aspects a sampling of RPV and its components is necessary to verify the theoretically calculated radiological data. The obtained results of activation and contamination are taken into account for the optimized dismantling and packaging strategy. The precise 3D-activation-model will reduce the necessary number and type of final disposal containers as security factors are minimized leading to a lower shielding effort, too. Besides, components or even parts of components may be subject of release measurement. In the end, costs can be reduced. In this context

  11. UDIN's dismantling projects

    International Nuclear Information System (INIS)

    Laffaille, C.

    1993-01-01

    The role of UDIN (Central unit for nuclear facility decommissioning) at the CEA is reviewed together with the main specific aspects of nuclear dismantling: the different options and dismantling strategies and costs. The characteristics of the main on-going projects are described: graphite-gas reactors (G2/G3), RAPSODIE (RNR), AT1 (pilot RNR fuel reprocessing plant), ELAN II B (Cesium source conditioning plant), EL4 (heavy water/CO2 reactor), RM2 (fuel control radio-metallurgical laboratory) and UB-UM (Uranium enrichment plant)

  12. Some aspects of reactor pressure vessel integrity

    International Nuclear Information System (INIS)

    Korosec, D.; Vojvodic, G.J.

    1996-01-01

    Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

  13. Assessment of the applicability of AWJ technique for dismantling the reactor of Fugen. Performance of underwater-cutting thick plate and testing of sound-based monitoring for underwater-cutting process

    International Nuclear Information System (INIS)

    Maruyama, Shin-ichiro; Nishio, Shin-ichi

    2010-01-01

    The reactor of Fugen is characterized by its double-walled pressure tube construction that is composed of pressure tubes and calandria tubes. The reactor dismantlement has been planning on dismantling it under water and the abrasive water jet (AWJ) underwater-cutting method is chosen as an option among simultaneous double tubes cutting technologies. For assessing the applicability of the AWJ cutting technology, a thick plate was cut under water by the small AWJ cutting machine. In addition, since cutting causes muddiness in water, cutting was monitored by the sound-based monitoring system which was adopted as a secondary cutting monitoring method. As a results, it was demonstrated that one-phase cutting was possible under water for a stainless-steel plate with 150mm thickness and that the relationship between cutting depth and capable cutting speed could be predictable. As for the sound-based cutting monitoring, the predictability whether or not cutting would be successful was verified by checking the change of sounds level. (author)

  14. Dismantling techniques

    International Nuclear Information System (INIS)

    Wiese, E.

    1998-01-01

    Most of the dismantling techniques used in a Decontamination and Dismantlement (D and D) project are taken from conventional demolition practices. Some modifications to the techniques are made to limit exposure to the workers or to lessen the spread of contamination to the work area. When working on a D and D project, it is best to keep the dismantling techniques and tools as simple as possible. The workers will be more efficient and safer using techniques that are familiar to them. Prior experience with the technique or use of mock-ups is the best way to keep workers safe and to keep the project on schedule

  15. Dismantling techniques

    Energy Technology Data Exchange (ETDEWEB)

    Wiese, E.

    1998-03-13

    Most of the dismantling techniques used in a Decontamination and Dismantlement (D and D) project are taken from conventional demolition practices. Some modifications to the techniques are made to limit exposure to the workers or to lessen the spread of contamination to the work area. When working on a D and D project, it is best to keep the dismantling techniques and tools as simple as possible. The workers will be more efficient and safer using techniques that are familiar to them. Prior experience with the technique or use of mock-ups is the best way to keep workers safe and to keep the project on schedule.

  16. Hot cell examination on the surveillance capsule of SA 533 cl. 1 reactor pressure vessel (1st test report)

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong Sun; Jung, Y. H.; Yoo, B. O.; Baik, S. J.; Oh, W. H.; Soong, W. S.; Hong, K. P

    2000-08-01

    The post-irradiated examinations such as impact test, tensile test, composition analysis and etc. were conducted to monitor and to evaluate the radiation-induced changes, so called radiation embrittlement, in the mechanical properties of ferritic materials. Those data should be applied to confirm safety as well as reliability of reactor pressure vessel. The scopes and contents of hot cell examination on the surveillance capsule are as follows; - Capsule transportation, cutting, dismantling and classification - Shim block and Dosimeter cutting and dismantling - Impact test - Tensile test - Composition analysis by EPMA - SEM observation on the fractured surface - Hardness test - Radwaste treatment.

  17. Further retardation could lead to a hold-up of nuclear reactor dismantling; Weitere Verzoegerungen koennten zu einem Stillstand des Kernkraft-Rueckbaus fuehren

    Energy Technology Data Exchange (ETDEWEB)

    Graf, Konstantin (comp.) [Innovations- und Technologieberatung Altran, Frankfurt am Main (Germany). Bereich Energy and Industry

    2015-07-01

    The following issues concerning the consequences of the German nuclear power phaseout are discussed: the cost of reactor dismantling could increase; the complete deconstruction of a nuclear power plant including environmental revitalization take a time of 10-15 years; the largest challenge is the still unsolved problem of final disposal; further retardations could trigger a complete deadlock of the deconstruction due to completely filled interim storage facilities. A further problem is the knowledge preservation due to the lack of students.

  18. Ultrasound periodic inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Haniger, L.

    1980-01-01

    Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

  19. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  20. Activation calculation for the dismantling and decommissioning of a light water reactor using MCNP™ with ADVANTG and ORIGEN-S

    Science.gov (United States)

    Schlömer, Luc; Phlippen, Peter-W.; Lukas, Bernard

    2017-09-01

    The decommissioning of a light water reactor (LWR), which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™) are applied and coupled with present-day activation and decay codes (ORIGEN-S). In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.

  1. Activation calculation for the dismantling and decommissioning of a light water reactor using MCNP™ with ADVANTG and ORIGEN-S

    Directory of Open Access Journals (Sweden)

    Schlömer Luc

    2017-01-01

    Full Text Available The decommissioning of a light water reactor (LWR, which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™ are applied and coupled with present-day activation and decay codes (ORIGEN-S. In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.

  2. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  3. Thorough Chemical Decontamination with the MEDOC Process : Batch Treatment of Dismantled Pieces or Loop Treatment of Large Components Such as the BR3 Steam Generator and Pressurizer

    International Nuclear Information System (INIS)

    Ponnet, M.; Klein, M.; Massaut, V.; Davain, H.; Aleton, G.

    2003-01-01

    The dismantling of the BR3-PWR reactor leads to the production of large masses of contaminated metallic pieces, including structural materials, primary pipings, tanks and heat exchangers. One of our main objectives is to demonstrate that we can minimize the volume of radioactive waste in an economical way, by the use of alternative waste routes, such as the clearance of materials after thorough decontamination. The SCKoCEN uses its own developed chemical decontamination process, so-called MEDOC (Metal Decontamination by Oxidation with Cerium), based on the use of cerium IV as strong oxidant in sulphuric acid with continuous regeneration using ozone. An industrial installation has been designed and constructed in close collaboration with Framatome-ANP (France). This installation started operation in September 1999 for the treatment of the metallic pieces arising from the dismantling of the BR3 reactor. Since then, more than 25 tons of contaminated material including primary pipes have been treated batchwise with success. 75 % of material could be directly cleared after treatment (Activity lower than 0.1 Bq/g for 60Co) and the other 25% free released after melting activity. The SCKoCEN performed in April 2002 the closed loop decontamination of the BR3 Steam Generator by connection of the MEDOC plant after few adaptations. The decontamination was done within 30 cycles in 3 weeks with consecutive steps like decontamination steps (injection of the solution into the SG) and regeneration steps with ozone. In total, 60 hours of decontamination at 70 C and 130 hours of regeneration were needed to reach the objectives. The tube bundle (600 m2) was attacked and about 10 (micro)m representing more than 41 kg of stainless steel and 2.06 GBq of 60Co was dissolved into the solution. The residual contamination measurements made directly into the water box are still going on, however it seems that the objective to reach the free release criteria after melting is achieved. The next

  4. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

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

  5. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Lee, A.D.

    1997-01-01

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  6. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

    1983-01-01

    This construction of a container, which is pressure-relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 900 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW)

  7. Reactor pressure vessel. Status report

    International Nuclear Information System (INIS)

    Elliot, B.J.; Hackett, E.M.; Lee, A.D.

    1996-10-01

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff's reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date

  8. Overview of recycling technologies for decommissioned materials. Lessons learned during the dismantling of a small PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Emond, O.; Ponnet, M.

    2001-01-01

    Full text: SCK CEN is dismantling its 11 MWe PWR reactor. The reactor was shutdown in 1987 after 25 years of operation and the dismantling started in 1990. For the management of the low radioactive materials, we apply a strategy promoting the minimisation of the production of radioactive waste and hence the maximisation of the production of recycled materials while keeping the costs as low as possible. The recycled materials are either reused in the non- nuclear industry as raw materials (metal scrap industry or building industry for the concrete) or recycled in the nuclear industry for specific applications (reuse of metals for fabrication of shielding, potential reuse of concrete for production of 'radioactive mortar'). The clearance of radioactive materials and their reuse require the strict respect of procedures and specifications. In our case, the Health Physics department under supervision of the Competent Authority establishes the procedures. This procedure is still a case by case practice but the legislation in Belgium is progressively put in place. For the recycling in the nuclear industry, we must respect the specifications of the end-user. Up to now, we have recycled low radioactive metals for the fabrication of shielding in the USA, so we had to respect the specifications of the melting facility and to obtain the authorisations for the transport abroad and for the transfer of property. Besides the radioactive waste route, we are using several evacuation routes for the dismantled materials: Evacuation of the cleared metals (iron, stainless steel, copper, electric motors...) to a local scrap dealer; Evacuation of metals to the Studsvik melting facility situated in Sweden: after clearance by the Swedish Authority, the non radioactive materials are sent to a local scrap dealer and the secondary radioactive waste is sent back to Belgium and conditioned by Belgoprocess. This technology further decontaminates the metals and allows performing an accurate

  9. Adaptation of a robot and tools for dismantling of a gas-cooled reactor

    International Nuclear Information System (INIS)

    Lewis, C.J.A.; Vibert, C.J.T.

    1989-01-01

    This report details the progress on a research programme to develop the techniques and design necessary to facilitate the use of commercially available industrial manipulator systems and cutting tools in nuclear environments, particularly that envisaged whilst decommissioning a gas-cooled reactor. The technology for the type of control and the machines to perform it already exist in the form of industrial-type robots. Development of the techniques for using these machines in a more operator-sensitive environment, together with the requirements of decontamination and radiation tolerance will enable them to be used in place of expensive purpose-built machines at a considerable cost saving. From this work it was possible to highlight the viability and associated costs of modifying a standard manipulator for use in decommissioning operations

  10. Nuclear installations: decommissioning and dismantling

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    This document is a compilation of seven talks given during the 1995 EUROFORUM conference about decommissioning and dismantling of Nuclear installations in the European Community. The first two papers give a detailed description of the legal, financial and regulatory framework of decommissioning and dismantling of nuclear facilities in the European Union and a review of the currently available decommissioning techniques for inventory, disassembly, decontamination, remote operations and management of wastes. Other papers describe some legal and technical aspects of reactor and plants dismantling in UK, Germany, Spain and France. (J.S.)

  11. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

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

  12. Results of reactor pressure vessels ISI

    International Nuclear Information System (INIS)

    Cepcek, S.

    1994-01-01

    To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

  13. Physics of pressurized water reactors

    International Nuclear Information System (INIS)

    Gruen, A.

    1980-01-01

    The objective of this lecture is to demonstrate typical problems and solutions encountered in the design and operation of PWR power plants. The examples selected for illustration refer to PWR's of KWU design and to results of KWU design methods. In order to understand the physics of a power reactor it is necessary to have some knowledge of the structure and design of the power plant system of which the reactor is a part. It is therefore assumed that the reader is familiar with the design of the more important components and systems of a PWR, such as fuel assemblies, control assemblies, core lay-out, reactor coolant system, instrumentation. (author)

  14. Dismantling of JPDR begins: to demonstrate advanced technology

    Energy Technology Data Exchange (ETDEWEB)

    1986-12-01

    The first dismantling of the Japan Power Demonstration Reactor (JPDR, BWR, 90 MWt, 12.5 MWe) began on December 4, 1986, claiming the attention of nuclear interests in Japan and overseas. The Japan Atomic Energy Research Institute undertook the project as the second phase of the six year program for dismantling the JPDR at the Tokai Research Establishment. It is the demonstration of the technology developed in the first phase of the program from 1981 to 1986, aiming at establishing a total system for dismantling commercial nuclear power plants in the furture. At the ceremony for the beginning of dismantling held on December 4 at the site, a special switch was operated to fire a gas burner, and cutting of the upper head of the reactor pressure vessel on the service floor of the reactor building began. The long term program on the development and utilization of nuclear energy in 1982 decided the basic policy on reactor decommissioning. Under this policy, in July, 1984, the nuclear subcommittee of the Advisory Committee for Energy set up the guideline for standardized decommissioning suitable to the actual situation in Japan. The schedule of the program, the development of eight fundamental techniques, disassembling techniques, decontamination, measurement and robotics are described. (Kako, I.).

  15. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  16. Dismantling system of concrete thermal shielding walls

    International Nuclear Information System (INIS)

    Machida, Nobuhiro; Saiki, Yoshikuni; Ono, Yorimasa; Tokioka, Masatake; Ogino, Nobuyuki.

    1985-01-01

    Purpose: To enable safety and efficient dismantling of concrete thermal shielding walls in nuclear reactors. Method: Concrete thermal shielding walls are cut and dismantled into dismantled blocks by a plasma cutting tool while sealing the top opening of bioshielding structures. The dismantled blocks are gripped and conveyed. The cutting tool is remote-handled while monitoring on a television receiver. Slugs and dusts produced by cutting are removed to recover. Since the dismantling work is carried out while sealing the working circumstance and by the remote control of the cutting tool, the operators' safety can be secured. Further, since the thermal sealing walls are cut and dismantled into blocks, dismantling work can be done efficiently. (Moriyama, K.)

  17. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

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

  18. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon

    2012-01-01

    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  19. A structural evaluation of the Shippingport reactor pressure vessel for transport impact conditions

    International Nuclear Information System (INIS)

    Witte, M.C.; Chou, C.K.

    1989-01-01

    The Shippingport Atomic Power Station in Shippingport, Pennsylvania, is being decommissioned and dismantled. This government-leased property will be returned, in a radiologically safe condition, to its owner. All radioactive material is being removed from the Shippingport Station and transported for burial to the DOE Hanford Reservation in Richland, Washington. The reactor pressure vessel (RPV) will be transported by barge to Hanford. This paper describes an evaluation of the structural response of the RPV to the normal and accident impact test conditions as required by the Code of Federal Regulations. 3 refs., 5 figs., 3 tabs

  20. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

    Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

  1. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  2. Brennilis, laboratory of dismantlement

    International Nuclear Information System (INIS)

    Dupin, L.

    2011-01-01

    This article comments some aspects of the dismantlement activity on the Brennilis site (in Brittany) where a heavy water reactor has been operated from 1966 to 1985. Half of the deconstruction work has been performed between 1996 and 2006. As the model proposed by EDF for this operation raised some questions, works have been stopped for a while, until July 2011 when a decree authorized them again, but for some parts of the site only. The reactor block must wait as no technical solution exists for storage. But, the experience from this site will be used for eight other first generation power plants

  3. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

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

  4. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  5. Towards EPR (European pressurized reactor)

    International Nuclear Information System (INIS)

    Anon.

    2003-01-01

    According to the French industry minister, it is nonsense continuing delaying the construction of an EPR prototype because France needs it in order to renew timely its park of nuclear reactors. The renewing is expected to begin in 2020 and will be assured with third generation reactors like EPR. A quick launching of the EPR prototype is necessary to have it being in service by 2012, the feedback operating experience that will be accumulated over the 8 years that will follow will be necessary to optimize the industrial version and to have it ready by 2020. The EPR reactor has indisputable assets: modern, safer, more competitive and it will produce less wastes than present nuclear reactors. The construction cost of an EPR prototype is estimated to 3 milliard Euros and the nuclear industry operators propose to finance it completely. The EPR prototype does not jeopardize the ambitious French program about renewable energy sources, France is committed to produce 21% of its electricity from renewable energies by 2010 and 10 milliard Euros will be invested over this period on wind energy. Nuclear energy and alternative energies must be considered as 2 aspects of a diversified energy policy. (A.C.)

  6. The good wealth of dismantlement

    International Nuclear Information System (INIS)

    Maincent, G.

    2009-01-01

    Civil engineering, mechanical and waste conditioning companies are working hard on the market of nuclear facilities dismantling. This market has a great future ahead of it in the ten years to come. According to the European Commission, 50 to 60 reactors among the 157 actually in service in the European Union should be dismantled by 2025. The cost per reactor is estimated to 10-15% of the initial investment, which represents an enormous amount of money, estimated to 20-39 billion euros for the only French nuclear park. In France, this market is shared by a core of about 20 companies, like Spie Nucleaire, Onet, Vinci (Nuvia) and Areva. Some dismantling sites require a specific skill, in particular those in relation with the research activity of the CEA (the French atomic energy commission) or involving specific technologies (research reactors, spent fuel reprocessing plants, sodium-cooled rectors..). (J.S.)

  7. Pressure releasing device for reactor container

    International Nuclear Information System (INIS)

    Takeda, Mika.

    1994-01-01

    In the present invention, dose rate to public caused by radioactive rare gases can be decreased. That is, a reactor container contains a reactor pressure vessel incorporating a reactor core. There are disposed a pressure releasing system for releasing the pressure in the reactor pressure vessel to the outside, and a burning device for burning gases released from the pressure releasing system. An exhaustion pipe is disposed to the pressure releasing system. A burning device is disposed to the exhaustion pipe. It is effective to dispose a ventilation port at a portion of the exhaustion pipe upstream of the burning device. In addition, the burning device may preferably be disposed in a multi-stage in the axial direction of the exhaustion pipe. With such procedures, hydrogen in gases discharged along with the release of the pressure in the container is burned. Buoyancy is caused to the exhaustion gases by heat energy upon burning. Since the exhaustion gases can reach a higher level by the buoyancy, the dose rate due to the rare gases can be reduced. (I.S.)

  8. N Reactor pressure tube 1350 postirradiation examination

    International Nuclear Information System (INIS)

    Cook, D.J.

    1977-01-01

    The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

  9. Method for dismantling shields

    International Nuclear Information System (INIS)

    Fukuzawa, Rokuro; Kondo, Nobuhiro; Kamiyama, Yoshinori; Kawasato, Ken; Hiraga, Tomoaki.

    1990-01-01

    The object of the present invention is to enable operators to dismantle shieldings contaminated by radioactivity easily and in a short period of time without danger of radiation exposure. A plurality of introduction pipes are embedded previously to the shielding walls of shielding members which contain a reactor core in a state where both ends of the introduction pipes are in communication with the outside. A wire saw is inserted into the introduction pipes to cut the shieldings upon dismantling. Then, shieldings can be dismantled easily in a short period of time with no radiation exposure to operator's. Further, according to the present invention, since the wire saw can be set easily and a large area can be cut at once, operation efficiency is improved. Further, since remote control is possible, cutting can be conducted in water and complicated places of the reactor. Biting upon starting the wire saw in the introduction pipe is reduced to facilitate startup for the rotation. (I.S.)

  10. The promising opportunity of dismantlement

    International Nuclear Information System (INIS)

    Anon.

    2009-01-01

    Civil engineering, mechanics and waste conditioning companies are thriving around the market of nuclear facilities dismantlement which is promised to a huge development in the coming decade. This paper presents a map of the opportunities of the dismantlement market throughout Europe (research and power reactors, fuel fabrication plants, spent fuel reprocessing plants) and a cost estimation of a given dismantling work with respect to the different steps of the work. In France a small core of about twenty companies is involved in nuclear dismantlement but the French market is also looking towards foreign specialists of this activity. The British market is also targeted by the French companies but for all the actors the technological or commercial advance gained today will be determining for the future markets. (J.S.)

  11. Dismantling of nuclear facilities

    International Nuclear Information System (INIS)

    Tallec, Michele; Kus, Jean-Pierre; Mogavero, Robert; Genelot, Gabriel

    2009-01-01

    Although the operational life of nuclear plants is long (around 60 years for French reactors) it is nonetheless limited in time, the stopping of it being essentially due to the obsolescence of materials and processes or to economic or safety considerations. The nuclear power plants are then subjected to cleanup and dismantling operations which have different objectives and require specific techniques. The cleanup and/or dismantling of a nuclear power produces significant quantities of waste which is generally of a different nature to that produced during the operation of the concerned plant. The radioactive waste produced by these operations is destined to be sent to the waste disposal facilities of the French National Agency for the Management of Nuclear Waste. (authors)

  12. The challenges of dismantling

    International Nuclear Information System (INIS)

    Sene, Monique; Lheureux, Yves; Leroyer, Veronique; Rollinger, Francois; Gauthier, Florence; Depauw, Denis; Reynal, Nathalie; Fraysse, Thierry; Burger, Eric; Bertrand, Adrien; Vallat, Christophe; Bernet, Philippe; Eimer, Michel; Boutin, Dominique; Bietrix, Philippe; Richard, Francoise; Piketty, Laurence; Mouchet, Chantal; Charre, Jean-Pierre

    2014-01-01

    This document gathers Power Point presentations which address the contexts and challenges of dismantling (legal framework, safety and radiation protection challenges, waste processing industry), and propose illustrations of dismantling challenges (example of operations to prepare EURODIF dismantling and CLIGEET work-group on EURODIF dismantling, examples of dismantling of EDF installations and CLIs' opinion on the dismantling of EDF installations, Brennilis dismantling follow-up performed by the CLI, examples of dismantling of CEA installations and opinion of a CLI on the dismantling of CEA installations)

  13. N Reactor pressure tube 2566 postirradiation examination

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

    Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

  14. The EL-4 reactor. Changing of a pressure tube on a test loop

    International Nuclear Information System (INIS)

    Foulquier, H.; Clara, P.

    1964-01-01

    Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

  15. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  16. Pressure test method for reactor pressure vessel in construction field

    International Nuclear Information System (INIS)

    Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

    1998-01-01

    Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

  17. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1992-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

  18. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1976-01-01

    Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

  19. Pressure suppression facility for reactor container

    International Nuclear Information System (INIS)

    Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

    1993-01-01

    In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

  20. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  1. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Margen, P H; Ahlstroem, P E; Pershagen, B

    1961-04-15

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

  2. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

    1961-04-01

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  3. Examination of VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Markulin, K.

    2008-01-01

    The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

  4. Cleansing and dismantling of CEA-Saclay nuclear licensed facilities

    International Nuclear Information System (INIS)

    Jeanjacques, Michel; Delaire, Isabelle; Glevarec, Rebecca; Mandard, Lionel; Martin, Jean-Louis; Serrano, Roger

    2013-01-01

    This summary presents the cleansing and dismantling operations currently realized on the CEA center of Saclay (CEA-Saclay). It was initiated at the beginning of the 2000 years a cleansing and dismantling program of the old Nuclear Licensed Facilities (NLF). Currently this program relates the dismantling operations to the Hot Laboratories (Laboratoires de Haute Activite: LHA) and the old workshops of the Liquid Waste Treatment Plant (Station des Effluents Liquides: STEL), the dismantling preparation of Ulysse reactor and the dismantling studies to the Solid Waste Management Plant (SWMP; Zone de Gestion des Dechets Solides) and the Osiris reactor. (authors)

  5. Leak detector for reactor pressure vessel

    International Nuclear Information System (INIS)

    Morimoto, Mikio.

    1991-01-01

    A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

  6. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  7. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

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

  8. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  9. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  10. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  11. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  12. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2008-04-01

    The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

  13. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  14. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  15. Radiation effects on reactor pressure vessel supports

    International Nuclear Information System (INIS)

    Johnson, R.E.

    1996-05-01

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

  16. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  17. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  18. Analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

    Oliveira, C.A.N. de; Augusto, O.B.

    1985-01-01

    This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt

  19. Pressure suppression system for a nuclear reactor

    International Nuclear Information System (INIS)

    Jost, N.

    1977-01-01

    The invention pertains to a pressure suppression system for PWR reactors where the parts enclosing the primary coolant are contained in two pressure-tight separate chambers. According to the invention, these chambers are partly filled with water and are connected with each other below the water surface. This way, gases cannot escape from the containment, not even if a valve and a line are damaged at the same time, as the vapours released condensate in the water of at least one of the other chambers. (HP) [de

  20. Head spray nozzle in reactor pressure vessel

    International Nuclear Information System (INIS)

    Hatano, Shun-ichi.

    1990-01-01

    In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

  1. Pressure thermal shock analysis for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

    2015-01-01

    The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

  2. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    Price, M.S.T.; Lafontaine, I.

    1985-01-01

    The purpose of this volume is to introduce the main types of nuclear reactor in the European Community (EC), select reference plants for further study, estimate the waste streams from the reference reactors, survey the transport regulations and assess existing containers

  3. Status of advanced small pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Peipei; Zhou Yun

    2012-01-01

    In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

  4. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  5. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  6. Seal analysis technology for reactor pressure vessel

    International Nuclear Information System (INIS)

    Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

    2009-01-01

    There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)

  7. Dismantling technologies trends

    International Nuclear Information System (INIS)

    Devaux, P.

    2009-01-01

    In this work dismantling technologies trends realized by the CEA are reviewed. There following technologies are presented: Data acquisition from facilities; Scenario studies; Remote handling and carriers; Dismantling techniques; Decontamination.

  8. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    Popp, P.

    1987-01-01

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  9. Pressure suppression device for nuclear reactor building

    International Nuclear Information System (INIS)

    Ikegame, Noboru.

    1992-01-01

    In a nuclear reactor building, there are disposed cooling coils connected to an air supply duct at the outside of the building, an air supply blower, an air supply duct having the top end opened, an exhaustion duct having the top end opened and a bypassing pipeline interposed between the exhaustion duct and the air supply duct on the side of the inlet of the cooling coils. In the reactor building, when a radioactive material leakage accident should occur, an isolation valve is closed to isolate the building from the outside. Further, bypassing isolation valve is opened to form a closed cooling circuit by the cooling coils, the air supply blower and the air supply duct, the exhaustion duct and the bypassing pipeline in the reactor building. With such a constitution, since air as the atmosphere in the building is circulated through the closed cooling circuit and cooled by the cooling coils, the temperature is not elevated. Accordingly, since the pressure elevation of the atmosphere in the building is suppressed, the atmosphere containing radioactive materials do not flow out of the building. (I.N.)

  10. EDF's dismantling experience

    International Nuclear Information System (INIS)

    Mira, J.J.

    1993-01-01

    The dismantling policy at EDF, taking into account technical, economical and socio-political factors, is presented. The various current realizations are reviewed and their dismantling solution discussed: Chinon A2, Chinon A1, Marcoule G1, G2, G3, Brennilis (EL4). Several dismantling projects are also described (Chinon A3, St-Laurent A1-A2, Chooz A). The various dismantling operations are presented and scheduled

  11. Nuclear reactor pressure vessel flaw distribution development

    International Nuclear Information System (INIS)

    Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

    1991-12-01

    Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

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

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

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

  13. Decree no. 2005-79 from January 26, 2005, authorizing the Atomic Energy Commission to proceed to the definitive shutdown and dismantling operations of the nuclear facility no.21, named Siloette research reactor, in the Grenoble city territory (Isere)

    International Nuclear Information System (INIS)

    2005-02-01

    On May 26, 2003, the French atomic energy commission (CEA) addressed an authorization demand for the definitive shutdown and dismantling of the Siloette research reactor. After a technical and administrative instruction of this demand by the French nuclear safety authority (ASN), a project of decree has been presented on July 6, 2004 at the permanent section of the inter-ministry commission of basic nuclear facilities. The commission gave its favourable judgment which is the object of this decree. (J.S.)

  14. Decree no. 2005-78 from January 26, 2005, authorizing the Atomic Energy Commission to proceed to the definitive shutdown and dismantling operations of the nuclear facility no.20, named Siloe reactor, in the Grenoble city territory (Isere)

    International Nuclear Information System (INIS)

    2005-02-01

    On March 19, 2003, the French atomic energy commission (CEA) addressed an authorization demand for the definitive shutdown and dismantling of the Siloe reactor. After a technical and administrative instruction of this demand by the French nuclear safety authority (ASN), a project of decree has been presented on July 6, 2004 at the permanent section of the inter-ministry commission of basic nuclear facilities. The commission gave its favourable judgment which is the object of this decree. (J.S.)

  15. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  16. Taking into account of dismantling constraints in the design of nuclear facilities

    International Nuclear Information System (INIS)

    Gouhier, E.; Moitrier, C.; Girones, P.; Pitrou, Y.; Poncet, P.; O'Sullivan, P.

    2014-01-01

    The taking into account of dismantling constraints in the design of nuclear facilities allows the reduction of the dosimetry during the dismantling operations, the reduction of the amount of wastes to manage and the saving of time and money by foreseeing an adequate and simple solution for each component. It is to notice that the strategy of life-extension strengthens that of dismantling because life-extension implies the possibility for any component of the reactor except the pressure vessel to be replaced. The feedback experience capitalized on various types of nuclear facilities have enabled IAEA and OECD to publish recommendations to facilitate dismantling. For instance, pipes and ventilation ducts must be designed to minimize the deposit of dust and residues, the natural porosity of concrete must be limited through the use of polishing products or a metal liner, the type and concentrations of impurities present in the structure materials must be controlled to limit radioactivation, the documentation describing the facility must be kept up to date, or the history of contamination events must be recorded all along the life of the facility. The integration of the dismantling constraints in the design stage is illustrated with 3 examples: the Georges Besse 2 enrichment fuel plant, new reactors (EPR, ASTRID and RJH), and ITER. (A.C.)

  17. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1986-09-01

    General descriptions of the main types of reactors in the European Economic Community are given, a series of reference plants selected for further study. Estimates are made of the radioactive decommissioning wastes for each, including neutron-activated and contaminated materials. Regulations governing the transport of radioactive materials, both international and national, are reviewed. (U.K.)

  18. Set-up of polarographic analytical methods in the framework of nuclear reactor dismantling en of the decontamination of metallic pieces

    International Nuclear Information System (INIS)

    Poirier, S.; Rahier, A.

    1996-06-01

    Differential pulse polarography has been used to measure several chemical species, relevant to the dismantling and the decontamination of a nuclear power reactor. First, a method which allows the determination of low concentrations of Co in stainless steels has been studied. Co 3+ is reduced in the presence of ethylenediamine at pH 7.5 at -0.47 V vs Ag/AgCl sat. A preliminary extraction of iron (and partially chromium) in diethylether is required. Interferences with iron and nickel have been completely eliminated without using any precipitation technique. Some complications may result from the precipitation of residual Cr 3+ in the presence of EDA, even when fluorides are added. Next, the measurements of the main components of steels have been carried out successfully. The reduction of CrO 4 2- is observed at -0.46 V vs Ag/AgCl sat. in a medium containing 0.1 M KOH, 0.5 M citric acid and 1 M NH 3 . Adding dimethylglyoxime in the same medium allows to identify the reduction to Fe 2+ and Ni 2+ respectively at -1.65 and -1.13 V vs Ag/AgCl sat. Finally, the reduction to Cr 3+ is observed at -1.2 V vs Ag/AgCl sat. in an acetic buffer containing 0.1 M EDTA

  19. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  20. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

    Biach, W.L.; Hill, R.; Hung, K.

    1992-01-01

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities

  1. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

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

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

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

  3. Dismantling of nuclear facilities

    International Nuclear Information System (INIS)

    Tallec, M.; Kus, J.P.

    2009-01-01

    Nuclear facilities have a long estimable lifetime but necessarily limited in time. At the end of their operation period, basic nuclear installations are the object of cleansing operations and transformations that will lead to their definitive decommissioning and then to their dismantling. Because each facility is somewhere unique, cleansing and dismantling require specific techniques. The dismantlement consists in the disassembly and disposing off of big equipments, in the elimination of radioactivity in all rooms of the facility, in the demolition of buildings and eventually in the reconversion of all or part of the facility. This article describes these different steps: 1 - dismantling strategy: main de-construction guidelines, expected final state; 2 - industries and sites: cleansing and dismantling at the CEA, EDF's sites under de-construction; 3 - de-construction: main steps, definitive shutdown, preparation of dismantling, electromechanical dismantling, cleansing/decommissioning, demolition, dismantling taken into account at the design stage, management of polluted soils; 4 - waste management: dismantlement wastes, national policy of radioactive waste management, management of dismantlement wastes; 5 - mastery of risks: risk analysis, conformability of risk management with reference documents, main risks encountered at de-construction works; 6 - regulatory procedures; 7 - international overview; 8 - conclusion. (J.S.)

  4. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  5. Study on treatment of dust by dismantling

    International Nuclear Information System (INIS)

    Torikai, K.; Suzuki, K.

    1987-01-01

    In dismantling of nuclear reactors, various kinds of treatment of dust generated by cutting or dismantling concrete structures of components of reactors are evaluated for safety, cost, and performance comparing the work in air with water. A method of dust treatment for work in air is discussed. The dry method has an easy operation in practice and a good performance in the equipment, but has problem on the prevention from radioactive contamination by diffusion of dust in air. For the purpose of advancing the strong points and eliminating the weak points in dry method, an improved venturi scrubber system is proposed for dismantling work as a dust collecting system. The system consists of dust absorbing pipe, dust collector, separator of dust and water and dust transfer equipment to a storage of waste. This system would be expected to have better performance and lower operating cost in decommissioning nuclear reactors, especially, the number of dust filters, for example, HEPA filters, will be considerably saved

  6. Prestressed concrete pressure vessels for nuclear reactors - 1973

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

  7. Cascading pressure reactor and method for solar-thermochemical reactions

    Science.gov (United States)

    Ermanoski, Ivan

    2017-11-14

    Reactors and methods for solar thermochemical reactions are disclosed. The reactors and methods include a cascade of reduction chambers at successively lower pressures that leads to over an order of magnitude pressure decrease compared to a single-chambered design. The resulting efficiency gains are substantial, and represent an important step toward practical and efficient solar fuel production on a large scale.

  8. Achievements and prospects of robotics in dismantling operations

    International Nuclear Information System (INIS)

    Clement, G.; Goetghebeur, S.; Ravera, J.P.

    1993-01-01

    After a definition of 'robotic systems' (poly functionality is the main concept), the nuclear facilities that have used robotic systems for their dismantling are reviewed; the various robot intervention domains in dismantling, the different types of machines and the work carried out by robots are presented. Difficulties arising from robot utilization for reactor dismantling, robot design considerations, reliability, personnel training needs, tooling and costs are discussed. Applicability criteria are derived concerning radio protection, hard working conditions, task complexity, multiplicity and quality, and costs

  9. Tools and tool application for the dismantling of the nuclear power plant Brennilis in France

    International Nuclear Information System (INIS)

    Bienia, Harald; Welbers, Philipp; Krueger, Peter; Noll, Thomas

    2012-01-01

    The EL-4 reactor in the NPP Brennilis in France is a CO2 cooled heavy water moderated test reactor with net power of 70 MW, the reactor started operation in 1967 and was decommissioned in 1985. Due to the construction features it was not necessary to enter the reactor area during operation, therefore the reactor pressure vessel and the surrounding piping systems are built in a very compact way. The dismantling procedures are therefore different from German BWR or PWR systems, the remote cutting and handling tools have to be adapted to the different features. Because of the high local dosage rate in the reactor hall it is also necessary to perform the erection of the dismantling equipment by robot systems. For cutting of the piping system a new plasma cutting technique, the hot wire method will be used. Other mechanical cutting techniques have to be used for instance for zircaloy containing components due to fire prevention purposes. The required time for tool and manipulator changes, including wearing part replacements constitute a significant part of the dismantling schedule. The suction/exhaust system for radioactive dust removal allowed a reduction of the total personal dose by one third of the allowed dose.

  10. Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

    International Nuclear Information System (INIS)

    Bodmann, E.

    1984-01-01

    Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

  11. SGN's Dismantling and Decommissioning engineering, projects experience and capabilities

    International Nuclear Information System (INIS)

    Destrait, L.

    1998-01-01

    Its experience in waste treatment, conditioning, storage and disposal, its cooperation with CEA and COGEMA Group in license agreements give SGN expertise in the decommissioning field. SGN's experience and background in all areas of nuclear facility decommissioning, such as chemical and mechanical cells, nuclear advanced reactors, reprocessing facilities result in fruitful references to the customers. The poster is presenting different achievements and projects with SGN's participation such as: - The decommissioning of Windscale Advanced Gas cooled Reactors (WAGR), in particular providing methodology and equipment to dismantle the Pressure and Insulation Vessel of the reactor. - The decommissioning plan of Ignalina (Lithuania) and Paldiski (Estonia), defining strategies, scenarios, necessary equipments and tools and choosing the best solutions to decommission the site under different influencing parameters such as cost, dose rate exposure, etc... - Th One Site Assistance Team (OSAT) at Chernobyl regarding the preparation works for the waste management and decommissioning of the plant. - The decommissioning of French nuclear facilities such as reprocessing (UP1) and reactor (EL4) plants. The important experience acquired during the facility management and during the first dismantling and decommissioning operations is an important factor for the smooth running of these techniques for the future. The challenge to come is to control all the operations, the choice of strategies, the waste management, the efficiency of tools and equipments, and to provide nuclear operators with a full range of proven techniques to optimise costs and minimize decommissioning personnel exposure. (Author)

  12. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  13. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  14. Midland reactor pressure vessel flaw distribution

    International Nuclear Information System (INIS)

    Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

    1993-12-01

    The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center's (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions

  15. Computerized cost model for pressurized water reactors

    International Nuclear Information System (INIS)

    Meneely, T.K.; Tabata, Hiroaki; Labourey, P.

    1999-01-01

    A computerized cost model has been developed in order to allow utility users to improve their familiarity with pressurized water reactor overnight capital costs and the various factors which influence them. This model organizes its cost data in the standard format of the Energy Economic Data Base (EEDB), and encapsulates simplified relationships between physical plant design information and capital cost information in a computer code. Model calculations are initiated from a base case, which was established using traditional cost calculation techniques. The user enters a set of plant design parameters, selected to allow consideration of plant models throughout the typical three- and four-loop PWR power range, and for plant sites in Japan, Europe, and the United States. Calculation of the new capital cost is then performed in a very brief time. The presentation of the program's output allows comparison of various cases with each other or with separately calculated baseline data. The user can start at a high level summary, and by selecting values of interest on a display grid show progressively more and more detailed information, including links to background information such as individual cost driver accounts and physical plant variables for each case. Graphical presentation of the comparison summaries is provided, and the numerical results may be exported to a spreadsheet for further processing. (author)

  16. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bergh, H.

    1987-01-01

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  17. Calandria cooling structure in pressure tube reactor

    International Nuclear Information System (INIS)

    Hyugaji, Takenori; Sasada, Yasuhiro.

    1976-01-01

    Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

  18. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  19. Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2016-01-01

    Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

  20. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  1. Pressurizer pump reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

  2. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  3. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  4. Dismantling of the 50 MW steam generator test facility

    International Nuclear Information System (INIS)

    Nakai, S.; Onojima, T.; Yamamoto, S.; Akai, M.; Isozaki, T.; Gunji, M.; Yatabe, T.

    1997-01-01

    We have been dismantling the 50MW Steam Generator Test Facility (50MWSGTF). The objectives of the dismantling are reuse of sodium components to a planned large scale thermal hydraulics sodium test facility and the material examination of component that have been operated for long time in sodium. The facility consisted of primary sodium loop with sodium heater by gas burner as heat source instead of reactor, secondary sodium loop with auxiliary cooling system (ACS) and water/steam system with steam temperature and pressure reducer instead of turbine. It simulated the 1 loop of the Monju cooling system. The rated power of the facility was 50MWt and it was about 1/5 of the Monju power plant. Several sodium removal methods are applied. As for the components to be dismantled such as piping, intermediate heat exchanger (IHX), air cooled heat exchangers (AC), sodium is removed by steam with nitrogen gas in the air or sodium is burned in the air. As for steam generators which material tests are planned, sodium is removed by steam injection with nitrogen gas to the steam generator. The steam generator vessel is filled with nitrogen and no air in the steam generator during sodium removal. As for sodium pumps, pump internal structure is pulled out from the casing and installed into the tank. After the installation, sodium is removed by the same method of steam generator. As for relatively small reuse components such as sodium valves, electromagnet flow meters (EMFs) etc., sodium is removed by alcohol process. (author)

  5. In-situ dismantling of plutonium-contaminated glove box

    International Nuclear Information System (INIS)

    Numata, Koji; Watanabe, Hisashi; Ishikawa, Hisashi; Miyo, Hiroaki; Ohtsuka, Katsuyuki

    1980-01-01

    A plutonium-contaminated glove box was dismantled along with the development of the treatment techniques for plutonium-bearing wastes. The objectives of this in-situ dismantling of the glove box are to reuse the Plutonium Fuel Fabrication Facility more efficiently, to reduce the volume of wastes generated during the dismantling, and to acquire dismantling techniques for decommissioning the Plutonium Fuel Fabrication Facility in the future. Prior to the dismantling works, a greenhouse for decontamination was installed, and the decontamination with surfactants was performed. Unremovable contamination was coated with paint. After this greenhouse was removed, the main greenhouse for dismantling and three greenhouses for contamination control were assembled. The main workers wearing protective devices engaged in dismantling works in the greenhouse. As the protective devices, anorak type PVC suits with air line masks, Howell type pressurized suits, and respirators were used. The tools used for the dismantling are a plasma cutter, an electric nibbler, an electric disk grinder, an electric circular saw and an electric jig saw. The results of the dismantling in-situ were compared with two previous cases of dismantling carried out by different procedures. In the case of in-situ dismantling, the volume of wastes was 1.6 - 1.8 m 3 /m 3 of glove box, and considerable reduction was realized. (Kako, I.)

  6. National School of Dismantling

    International Nuclear Information System (INIS)

    Ivaldi, Fabienne

    2003-01-01

    The National Institut of Nuclear Sciences and Techniques founded of 2001 a National School of Dismantling, NSD, at the end, which was validated by CEA, COGEMA, EDF and ANDRA. This school addresses four major issues: Decontamination; Dismantling; Demolition and waste Disposal (4D). Dedicated for instructing scientific and technical knowledge and know-how, needed in dismantling the nuclear installations, NSD has as targets: - personnel at engineering and operational level; - personnel occupied with involved trades from conception through intervention; - students and employees on leave; - employees while training on the job. Initial basic education for students in collaboration with schools and universities concerns: - master degree in radioactive waste management; - master degree in dismantling; - professional license in 3 D; - pro 4 D graduation. NSD is also engaged in continual formation for employees qualified, or not, adapted to the needs generated by the following tasks and personnel: - introduction in dismantling; - project team; - specialist engineer; - team head; - agent for remedial action; - agent for dismantling. The National School of Dismantling joins a network of human and technological capabilities confined within the 4 D frame, namely: - scientific and technical competencies (experts, instructors working in the nuclear field and dismantling); - pedagogical competence (professionals from basic and continual education); - specific material means such as those used by construction site schools, mock-ups, rooms for practical training etc

  7. Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Bros, J.; Ballesteros, A.; Lopez, A.

    1993-01-01

    Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

  8. Decommissioning experience of the Japan power demonstration reactor

    International Nuclear Information System (INIS)

    Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

    1992-01-01

    Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

  9. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Prodea, Iosif; Catana, Alexandru

    2010-01-01

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  10. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    Kempf, Rodolfo; Fortis, Ana M.

    2007-01-01

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

  11. Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Najzer, M.

    1982-01-01

    Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

  12. TORT application in reactor pressure vessel neutron flux calculations

    International Nuclear Information System (INIS)

    Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

    1994-01-01

    The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

  13. 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.

  14. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-10

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

  15. Abrasive water jet cutting technique for biological shield concrete dismantlement

    International Nuclear Information System (INIS)

    Konno, T.; Narazaki, T.; Yokota, M.; Yoshida, H.; Miura, M.; Miyazaki, Y.

    1987-01-01

    The Japan Atomic Energy Research Institute (JAERI) is developing the abrasive-water jet cutting system to be applied to dismantling the biological shield walls of the JPDR as a part of the reactor dismantling technology development project. This is a total system for dismantling highly activated concrete. The concrete biological shield wall is cut into blocks by driving the abrasive-water jet nozzle, which is operated with a remote, automated control system. In this system, the concrete blocks are removed to a container, while the slurry and dust/mist which are generated during cutting are collected and treated, both automatically. It is a very practical method and will quite probably by used for actual dismantling of commercial power reactors in the future because it can minimize workers' exposure to radioactivity during dismantling, contributes to preventing diffusion of radiation, and reduces the volume of contaminated secondary waste

  16. Innovative inspection system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Mertens, K.; Trautmann, H.

    1999-01-01

    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [de

  17. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  18. Excess-pressure suppression device in a reactor container

    International Nuclear Information System (INIS)

    Nishio, Masahide

    1985-01-01

    Purpose: To reliably decrease the radioactivity of radioactive gases when they are released externally. Constitution: The exit of a gas exhaust pipe for discharging gases in a reactor container, on generation of an excess pressure in the reactor container upon loss of coolant accident, is adapted to be always fluided in the cooling tank. Then, the exhaust gases discharged in the cooling tank is realeased to the atmosphere. In this way, the excess pressure in the reactor container can be prevented previously and the radioactivity of the gases released externally is significantly reduced by the scrubbing effect. (Kamimura, M.)

  19. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  20. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  1. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  2. Some regulation aspects in dismantling

    International Nuclear Information System (INIS)

    Niel, J.C.

    1993-01-01

    In the French regulation framework, operations linked to dismantling are controlled by an overall technical and legislative system applied to all the different stages of the facility (commissioning, etc.). Government control on facilities under dismantling is aimed at dismantling operation safety and security, and dismantling waste processing in order to ensure public and environmental protection

  3. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  4. Plastic reactor suitable for high pressure and supercritical fluid electrochemistry

    DEFF Research Database (Denmark)

    Branch, Jack; Alibouri, Mehrdad; Cook, David A.

    2017-01-01

    The paper describes a reactor suitable for high pressure, particularly supercritical fluid, electrochemistry and electrodeposition at pressures up to 30 MPa at 115◦C. The reactor incorporates two key, new design concepts; a plastic reactor vessel and the use of o-ring sealed brittle electrodes...... by the deposition of Bi. The application of the reactor to the production of nanostructures is demonstrated by the electrodeposition of ∼80 nm diameter Te nanowires into an anodic alumina on silicon template. Key advantages of the new reactor design include reduction of the number of wetted materials, particularly...... glues used for insulating electrodes, compatability with reagents incompatible with steel, compatability with microfabricated planar multiple electrodes, small volume which brings safety advantages and reduced reagent useage, and a significant reduction in experimental time....

  5. Use of superheated steam to anneal the reactor pressure vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1994-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 degrees F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 degrees F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors

  6. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

  7. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  8. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

  9. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  10. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

    Siler, W.C.; Marda, R.S.; Smith, W.R.

    1977-01-01

    Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

  11. Pressurized heavy-water reactor safety

    International Nuclear Information System (INIS)

    Pease, L.; Wilson, R.

    1977-09-01

    CANDU-PWR type reactors routinely release small amounts of radioactive liquids and gases and large quantities of low-grade waste heat. Radioactive emissions are usually below 1% of the derived release limits based on ICRP limits. Waste heat is common to all power plants and is not foreseen as a problem in Canadian conditions. Risk analysis shows a very low accident probability for CANDU type reactors. Multiple barriers to release of radionuclides, quality assurance, control, and inspection, containment systems, the shutdown system, the ECCS, and leak-before-break design, would all combine to mitigate the effects of an accident. (E.C.B.)

  12. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  13. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    Leroy, C.; Bitsch, D.; Millot, J.P.

    1985-10-01

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  14. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  15. Electrochemical noise measurements under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2000-01-01

    Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

  16. SCW Pressure-Channel Nuclear Reactor Some Design Features

    Science.gov (United States)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  17. Pressure equalization systems in pressurized water reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.; Wunderlich, F.

    1979-01-01

    For the development of a pressure reduction system in PWR fuel rods the capability of charcoal to adsorb Helium, Xenon and Krypton at temperatures of about 300 0 C was investigated. The influence of the adsorption on fuel rod internal pressure and in creep strain on the tube was evaluated in a design study. (orig.) [de

  18. Dismantling of nuclear facilities: the industrial know-how

    International Nuclear Information System (INIS)

    Lellament, R.

    2004-01-01

    Numerous nuclear facilities in laboratories or research reactors have been decommissioned and dismantled over the 2 last decades throughout the world. The valuable feedback experience has allowed nuclear industry to design, upgrade and test specific techniques for dismantling. These techniques are efficient although they have been validated on a reduced number of nuclear power plants. In France only 3 power units have been dismantled: Chinon A1, A2 and Brennilis (EL4) and they are not representative of the real park of EDF'reactors. 6 PWR-type reactors have already been dismantled in the Usa. The results of a survey concerning 26 countries shows that the dismantling cost is around 320 dollars/kWe, it represents 15% of the construction cost which is far from being excessive as it is often read in the media. The dismantling costs can be broken into: - de-construction (25-55%), - wastes from dismantling (17-43%), - security and monitoring (8-13%), - site reclamation (5-13%), and - engineering and project management (5-24%). (A.C.)

  19. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

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

  20. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  1. Assessment of a small pressurized water reactor for industrial energy

    International Nuclear Information System (INIS)

    Klepper, O.H.; Fuller, L.C.; Myers, M.L.

    1977-01-01

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton

  2. Enriched uranium cycles in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Mazzola, A.

    1994-01-01

    A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

  3. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  4. Thermal annealing of an embrittled reactor pressure vessel

    International Nuclear Information System (INIS)

    Mager, T.R.; Dragunov, Y.G.; Leitz, C.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

  5. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  6. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-08

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

  7. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  8. Decay ratio estimation in pressurized water reactor

    International Nuclear Information System (INIS)

    Por, G.; Runkel, J.

    1990-11-01

    The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

  9. Assessment of the integrity of WWER type reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1995-01-01

    Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

  10. Acoustic Emission for on-line reactor pressure boundary monitoring

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

    1985-01-01

    The program objective is to develop AE for continuous surveillance to assess flaw growth in reactor pressure boundaries. Technology in the laboratory is being evaluated on structures. Results have demonstrated basic feasibility of the program objective. AE monitoring a long term fatigue test of a pressure vessel demonstrated an instrument system, and the ability to detect unexpected as well as well as known fatigue cracks. Monitoring a nuclear reactor system shows that the coolant flow noise problem is manageable and AE can be detected under simulated operating conditions

  11. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  12. Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

    International Nuclear Information System (INIS)

    Adamec, P.

    2000-12-01

    Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

  13. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  14. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  15. In-service ultrasonic inspection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Prepechal, J.; Sulc, J.

    1982-01-01

    Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

  16. Absorber rod bundle actuator in a pressurized water nuclear reactor

    International Nuclear Information System (INIS)

    Martin, J.; Peletan, R.

    1984-01-01

    The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

  17. High pressure sealing systems for nuclear reactors

    International Nuclear Information System (INIS)

    Garam, E. de

    1993-01-01

    TIA is the FRAMATOME Division in charge of design, manufacture maintenance and improvement of reactor core instrumentation. In the course of its activities, TIA was rapidly confronted with problems of leakage occurring in PWR in-core instrumentation, both in the neutron flux measurement system (flux thimbles and thimble guide tubes) and in the equipment used for core temperature sensing. TIA has likewise placed emphasis, in setting objectives for its operations, on improving instrumentation reliability, reducing maintenance costs and limiting the radiation doses sustained during maintenance. The very satisfactory results achieved by TIA in all of these areas have led us to look to the future with confidence. The purpose of this presentation is to describe the various improvements devised by TIA over the years and to take inventory of the experience gained by the Division with instrumentation for all types of nuclear power plants. (author)

  18. A powerful methodology for reactor vessel pressurized thermal shock analysis

    International Nuclear Information System (INIS)

    Boucau, J.; Mager, T.

    1994-01-01

    The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

  19. The dismantling of nuclear installations

    International Nuclear Information System (INIS)

    Lacoste, A.C.; Duthe, M.; Mignon, H.; Lambert, F.; Pradel, Ph.; Hillewaere, J.P.; Dupre la Tour, St.; Mandil, C.; Weil, L.; Eickelpasch, N.; Finsterwalder, L.

    1997-01-01

    for nuclear installations, the dismantling is an important part of their exploitation. The technology of dismantling is existing and to get a benefit from the radioactive decay, it seems more easy for operating company such E.D.F. to wait for fifty years before dismantling. But in order to get the knowledge of this operation, the Safety Authority wanted to devote this issue of 'Controle'to the dismantling method. This issue includes: the legal aspects, the risks assessment, the dismantling policy at E.D.F., the site of Brennilis (first French experience of dismantling), the dismantling techniques, the first dismantling of a fuel reprocessing plant, comparison with classical installations, economic aspect, some German experiences, the cleansing of the american site of Handford. (N.C.)

  20. Pressure suppression device for a reactor container

    International Nuclear Information System (INIS)

    Shimizu, Toshiaki

    1982-01-01

    Purpose: To prevent damages in drain pipes or the likes upon the water level increase due to blowing of incompressible gases. Constitution: An exhaust pipe for guiding escaping steams is connected to a main steam releaf valve. The exhaust pipe is guided into pressure-suppression-chamber water through the inside of a dry-well and by way of a vent pipe, a vent header and a drain pipe or a downcomer. Since the exhaust pipe is not exposed to the water surface inside the pressure suppression chamber, even if steams blow out into the dry-well by the rapture of pipeways or the likes to rapidly increase the water level, the water surface does not hit on the exhaust pipe, whereby the damages for the exhaust pipe and support members can be prevented to improve the reliability. (Seki, T.)

  1. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

    Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

    1984-01-01

    The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

  2. Pressure vessel codes: Their application to nuclear reactor systems

    International Nuclear Information System (INIS)

    1966-01-01

    A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

  3. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  4. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  5. Marine reactor pressure vessels dumped in the Kara Sea

    International Nuclear Information System (INIS)

    Mount, M.E.

    1997-01-01

    Between 1965 and 1988, 16 marine reactors from seven Russian submarines and the icebreaker Lenin, each of which suffered some form of reactor accident, were dumped in a variety of containments, using a number of sealing methods, at five sites in the Kara Sea. All reactors were dumped at sites that varied in depth from 12 to 300 m and six contained their spent nuclear fuel (SNF). This paper examines the breakdown of the reactor pressure vessel (RPV) barriers due to corrosion, with specific emphasis on those RPVs containing SNF. Included are discussions of the structural aspects of the steam generating installations and their associated RPVs, a summary of the disposal operations, assumptions on corrosion rates of structural and filler materials, and an estimate of the structural integrity of the RPVs at the present time (1996) and in the year 2015

  6. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  7. Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-10-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

  8. Leak detection device for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Ikeda, Jun.

    1988-01-01

    Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

  9. Dual pressurized light water reactor producing 2000 M We

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

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

  10. Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

    2009-01-01

    An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

  11. Evaluation formulas of manpower needs for dismantling of equipment in FUGEN-3. Dismantling process of the condenser removal

    International Nuclear Information System (INIS)

    Kubota, Shintaro; Izumo, Sari; Usui, Hideo; Kawagoshi, Hiroshi; Koda, Yuya; Nanko, Takashi

    2014-07-01

    Japan Atomic Energy Agency (JAEA) has been developing the PRODIA code which supports to make decommissioning plan and has been preparing evaluation formulas. Manpower needs for the dismantling of the condenser that had conducted from 2010 to 2012 was analyzed and compared with existing evaluation formulas. Applicability of evaluation formulas for a large scale reactor facility was confirmed in dismantling of the heat insulating materials and reliability of unit productivity factor was improved. The evaluation formula of work for clearance was made in dismantling of pipes and supports. Unit productivity factor of dismantling of feed water heaters which is applicable for a large scale reactor facility was derived. For derivation of unit productivity factor, statistically meaningful data was provided from the dismantling of the condenser. Manpower needs for dismantling of the condenser has positive correlation to the weight of equipment and can be described in linear expression. Reliability of each unit productivity factor will be improved with accumulating actual dismantling data in future. (author)

  12. LEP dismantling starts

    CERN Multimedia

    2000-01-01

    Since the end of November, various teams have been getting stuck into dismantling the LEP accelerator and its four experiments. After making the installations safe, the dismantling and removal of 40,000 tonnes of equipment is underway. Down in the tunnel, it is a solemn moment. It is 10 o'clock on 13 December and Daniel Regin, one of those heading the dismantling work, moves in on a magnet, armed with a hydraulic machine. Surrounded by teams gathered there for a course in dismantling, he makes the first cut into LEP. The great deconstruction has begun. In little over than a year, the accelerator will have been cleared away to make room for its successor, the LHC. The start of the operation goes back to 27 November. Because before setting about the machine with hydraulic shears and monkey wrenches, LEP had first to be made safe - it was important to make sure the machine could be taken apart without risk. All the SPS beam injection systems to LEP were cut off. The fluids used for cooling the magnets and superc...

  13. Advanced Approach of Reactor Pressure Vessel In-service Inspection

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Pajnic, M.

    2006-01-01

    The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

  14. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  15. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  16. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  17. Manipulator for testing a top-opened reactor pressure vessel

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1991-01-01

    The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

  18. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  19. Dismantling of nuclear facilities and related problems - Conference proceedings

    International Nuclear Information System (INIS)

    Tournebize, Frederic; Bordet, Didier; Charlety, Philippe; Gore, Thierry; Estrade, Jerome; Lemaire, Hermine; Ginet, Annick; Fabrier, Lionel; Evrard, Lydie; Furois, Timothee; Butez, Marc; Dutzer, Michel; Faure, Vincent; Billarand, Yann; Menuet, Lise; Lahaye, Thierry; Pin, Alain; Mougnard, Philippe; Charavy, Sylvain; Poncet, Philippe; Moggia, Fabrice; Dochy, Arnaud; Benjamin, Patrick; Poncet, Pierre-Emmanuel; Beneteau, Yannick; Richard, Jean-Baptiste; Pellenz, Gilles; Ollivier Dehaye, Catherine; Gerard, Stephane; Denissen, Luc; Davain, Henri; Duveau, Florent; Guyot, Jean-Luc; Ardellier, Luc

    2012-11-01

    The oldest French nuclear facilities, built for some of them in the 1950's for research or power generation purposes, have reached more or less the end of their life. More than 30 facilities have entered the shutdown or dismantling phase, among which 8 reactors of the very first generations of Electricite de France (EdF) reactors. The aim of this two-days conference is to take stock of the present day status and perspectives of the dismantling activity, to approach the question of the management of the wastes produced, and to share experience about large scale operations already carried out. This document gathers the available presentations given during this conference: 1 - the 'Passage' project (F. Tournebize); 2 - CEA-Grenoble: from Louis Neel to key enabling technologies (D. Bordet); 3 - Dismantling actions in France (L. Evrard); 4 - Securing control of long-term charges funding (T. Furois); 5 - Waste disposal projects and their contribution to the management of dismantling wastes (M. Butez); 6 - Specificities linked with dismantling activities (Y. Billarand); 7 - Dismantling safety: the ASN's point of view (L. Evrad); 8 - Labor Ministry viewpoint about the dismantling related questions (T. Lahaye); 9 - Consideration of organizational and human factors in dismantling operations: a new deal in the operators-service providers relation (L. Menuet); 10 - Diploma and training experience (A. Pin); 11 - Glove-boxes dismantling at La Hague plant - status and experience feedback (P. Mougnard); 12 - Dismantling of Siloe reactor (CEA-Grenoble): application of the ALARA approach (P. Charlety); 13 - BR3 - a complex dismantling: the neutron shield tank (NST) in remote operation and indirect vision (L. Denissen); 14 - Cleansing and dismantling of the Phebus PF containment (S. Charavy); 15 - Integration of dismantling at the design and exploitation stages of nuclear facilities (P. Poncet); 16 - Consideration during the design and exploitation stages of dispositions aiming at

  20. Nuclear power - replacement of pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

  1. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

  2. Status of the Digital Mock-up System for the dismantling of the nuclear facilities

    International Nuclear Information System (INIS)

    Park, Hee Seoung; Kim, S. K.; Lee, K. W.; Oh, W. J.

    2004-12-01

    The database system have already developed is impossible to solve a quantitative evaluation about a various situation from the dismantle activities of the reactor had contaminated with radioactivity. To satisfy the requirements for safety and economical efficiency among a major decommissioning technologies, it need a system that can evaluate and estimate dismantling scheduling, amount of radioactive waste being dismantled, and decommissioning cost. We have review and analyzed status of the digital mock-up system to get a technical guide because we have no experience establishment of one relation to dismantling of research reactor and nuclear power plant

  3. Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

    International Nuclear Information System (INIS)

    Jooho, W.; Baldwin, G.T.

    2005-01-01

    One critical aspect of any denuclearization of the Democratic People's Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for 'complete, verifiable and irreversible dismantlement,' or 'CVID.' It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long completion times

  4. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  5. Sampling of Reactor Pressure Vessel and Core Internals

    International Nuclear Information System (INIS)

    Oberhaeuser, R.

    2011-01-01

    Decommissioning and dismantling of nuclear power plants is a growing business, as a huge number of plants built in the 1970s have now reached their lifetime. It is well known that dismantling a nuclear power plant means an extraordinary expense for the owner respectively operator. Besides the dismantling works, the disposal of activated components and other nuclear waste is very expensive. Moreover, the fact that, in most countries, final disposal facilities are not available yet implies the need for interim storage on-site in specially built facilities. It can be concluded that a special attention is paid on producing a minimal radioactive waste volume. For this, optimized dismantling and packaging concepts have to be developed. The challenge is a fair balance between the obtainment of optimized packing and on the other side the fulfillment of stringent regulations set by the authorities and the storage requirements. The basis of a well-founded, optimized dismantling and packaging concept must always be the detailed knowledge of the radiological condition of the component to be dismantled. In the best case a 3- dimensional activation model contributes to this basis.

  6. Neutron physics of a high converting advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Berger, H.D.

    1985-01-01

    The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de

  7. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  8. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    1996-01-01

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  9. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  10. State of the art of the advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Chawla, R.

    1987-01-01

    A review is given of the present status of the works concerned with an advanced pressurized water reactor (APWR). It includes the following items: reactor physics, thermal and hydraulic investigations and other engineering aspects as well as an analysis of electricity generation cost and long-term problems of embedding the APWR in a plutonium economy. As a summary it can be stated that there are discernible no principal obstacles of technically accomplishing an APWR, but there will be necessary considerable expenses in research and development works if it should be intended to start commercial service of an APWR up to the end of this century. (author)

  11. Calculation of external exposure during transport and disposal of radioactive waste arisen from dismantling of steam generator

    International Nuclear Information System (INIS)

    Hornacek, M.; Necas, V.

    2014-01-01

    The dismantling of large components (reactor pressure vessel, reactor internals, steam generator) represents complex of processes involving preparation, dismantling, waste treatment and conditioning, transport and final disposal. To optimise all of these activities in accordance with the ALARA principle the prediction of the exposure of workers is an essential prerequisite. The paper deals with the calculation of external exposure of workers during transport and final disposal of heat exchange tubes of steam generator used in Slovak nuclear power plant V1 in Jaslovske Bohunice. The type of waste packages, the calculation models of truck and National Radioactive Waste Repository in Mochovce are presented. The detailed methodology of radioactive waste disposal is showed and the degree of influence of time decay (0, 5 and 10 years) on the radiological conditions during transport and disposal is studied. All of the results do not exceed the limits given in Slovak and international regulatory documents. (authors)

  12. Dismantling large components at the Jose-Cabrera NPP (CNJC) in Spain

    International Nuclear Information System (INIS)

    Santiago, Juan Luis

    2012-01-01

    Located in central Spain, near Madrid, the Jose-Cabrera NPP (also known as Zorita) is the first PWR to be dismantled in Spain. The unit is a one-loop Westinghouse PWR, with a capacity of 150 MW. The plant was shut down in 1996 and ENRESA (Empresa Nacional de Residuos Radioactivos) has decided its prompt decommissioning, starting in 2010. In preparation for decommissioning, a full system decontamination (FSD) of the whole reactor cooling system (including the reactor vessel in the flow path) was carried out in 2006-7. The large components to be dismantled include: the reactor pressure vessel (RPV) and the internals; the vessel head; the SG; the pressurizer and the surge line; the reactor coolant pump, and the primary loop piping The objective of the project is not only to ensure the safe and efficient dismantling of those large components, but also to gain experience and to learn lessons to be applied during the future decommissioning and dismantling of the remaining six operating PWRs in Spain, whose operational lives are currently planned to end between 2021 and 2028. ENRESA has defined a waste-management policy for decommissioning activities, which includes Waste-management routes and optimisation. A case study describes the results obtained by ENRESA in the specific case of CNJC large components (including the RPV) dismantling project: Removal and conditioning of large components as a single piece is not considered a viable option. Segmentation therefore is required and 2 options have been analysed: large pieces for disposal in a large container and small pieces for disposal in approved concrete packages (CE-2a and the smaller CE-2b). The use of the CE-2b package is a feasible option and is easy to implement as a logical extension from the CE-2a. The use of the CE-2b package results in an important reduction in the total volume of final waste packages and does not require, in itself, any changes in the current waste handling and kinematics. The large size

  13. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  14. Superheated steam annealing of pressurized water reactor vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1993-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  16. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  17. Dismantling of Vandellos I

    International Nuclear Information System (INIS)

    Armada, J. R.

    2003-01-01

    Spain is witnessing the phase-out of a nuclear power plant. It is a unique experience in our country and therefore the dismantling work has been watched closely, not only from here but also from abroad. The Empresa Nacional de Residuos Radiactivos (ENRESA) is in charge of managing the dismantling and decommissioning work of the Vandellos-I nuclear power plant, located in the municipality of L'Hospitalet de l'Infant (Tarragona). the work began five years ago and has been executed on schedule. the appearance of what was one of the first Spanish commercial nuclear power plants has been changed radically to leave premises suitable for any other activity. (Author)

  18. EDF's nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1987-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

  19. Needs for evaluated covariance data for reactor pressure vessel dosimetry

    International Nuclear Information System (INIS)

    Maerker, R.E.; Broadhead, B.L.; Wagschal, J.J.

    1992-01-01

    This report discusses new methodology for quantifying and then reducing uncertainties in the calculated pressure vessel fluences of a pressurized water reactor (PWR). The technique involves combining the integral results of the calculated and measured PWR surveillance dosimetry activities with the differential data used in the calculations, along with covariances of all the quantities, into a generalized linear least-squares adjustment procedure. Based on analysis of both PWRs and test reactor benchmarks, substantial evidence now exists to support the conclusion that, of all the nuclear as well as non-nuclear differential data considered, ENDF/B-VI values of the total inelastic iron cross sections and their covariances are the most important data controlling the outcome of the adjustment procedure. Predicted adjustments in these cross sections provided the stimulus for new measurements, the results of which impacted the ENDF/B-VI evaluation of iron 56

  20. Development of pressure boundaries leak detection technology for nuclear reactor

    International Nuclear Information System (INIS)

    Zhang Yao; Zhang Dafa; Chen Dengke; Zhang Liming

    2008-01-01

    The leak detection for the pressure boundaries is an important safeguard in nuclear reactor operation. In the paper, the status and the characters on the development of the pressure boundaries leak detection technology for the nuclear reactor were reviewed, especially, and the advance of the radiation leak detection technology and the acoustic emission leak detection technology were analyzed. The new advance trend of the leak detection technology was primarily explored. According to the analysis results, it is point out that the advancing target of the leak detection technology is to enhance its response speed, sensitivity, and reliability, and to provide effective information for operator and decision-maker. The realization of the global leak detection and the whole life cycle health monitoring for the nuclear boundaries is a significant advancing tendency of the leak detection technology. (authors)

  1. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  2. 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

  3. Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.

    1988-05-01

    The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.

  4. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  5. Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Zahoor, A.

    1988-01-01

    The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates. (orig.)

  6. EDF'S nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1988-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

  7. Thermohydraulic feedbacks in self-pressurized reactor systems

    International Nuclear Information System (INIS)

    Fiebig, R.

    1977-01-01

    The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controlable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

  8. Thermohydraulic feedbacks in self-pressurized reactor systems

    International Nuclear Information System (INIS)

    Fiebig, R.

    1977-01-01

    The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controllable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

  9. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  10. Welding in repair of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Pilous, V.; Kovarik, R.

    1987-01-01

    Specific welding conditions are described in repair of the pressure vessels of nuclear reactors in operation and the effect is pointed out to of neutrons on changes in steel properties. Some of the special regulations are discussed to be observed in welding jobs. The welding methods are briefly described; the half-bead method is most frequently used. It is stressed that the defect must first be identified using a nondestructive method and the stages must be defined of the welding repair of the pressure vessel. (J.B.). 4 figs., 1 tab., 16 refs

  11. Annealing of the BR3 reactor pressure vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

    1985-01-01

    The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

  12. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  13. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  14. Remote tool development for nuclear dismantling operations

    International Nuclear Information System (INIS)

    Craig, G.; Ferlay, J.C.; Ieracitano, F.

    2003-01-01

    Remote tool systems to undertake nuclear dismantling operations require careful design and development not only to perform their given duty but to perform it safely within the constraints imposed by harsh environmental conditions. Framatome ANP NUCLEAR SERVICES has for a long time developed and qualified equipment to undertake specific maintenance operations of nuclear reactors. The tool development methodology from this activity has since been adapted to resolve some very challenging reactor dismantling operations which are demonstrated in this paper. Each nuclear decommissioning project is a unique case, technical characterisation data is generally incomplete. The development of the dismantling methodology and associated equipment is by and large an iterative process combining design and simulation with feasibility and validation testing. The first stage of the development process involves feasibility testing of industrial tools and examining adaptations necessary to control and deploy the tool remotely with respect to the chosen methodology and environmental constraints. This results in a prototype tool and deployment system to validate the basic process. The second stage involves detailed design which integrates any remaining technical and environmental constraints. At the end of this stage, tools and deployment systems, operators and operating procedures are qualified on full scale mock ups. (authors)

  15. Technology and costs for dismantling a Swedish nuclear power plant

    International Nuclear Information System (INIS)

    1979-10-01

    Various estimates concerning the costs of decommissioning a redundant nuclear power reactor to the green fields state are given in the literature. The purpose of this study is to provide background material for the Swedish nuclear power utilities to estimate the costs and time required to dismantle an ASEA-ATOM Boiling Water Reactor. The units Oskarshamn II and Barsebeck 1, both with an installed capacity of approximately 600 MW, serve as reference plants. The time of operation before final shutdown is assumed to be 40 years. Dismantling operations are initiated one year after shutdown. When the dismantling of the plant is finished, the site is to be released for unrestricted use. The costs for dismantling and subsequent final disposal of the radioactive waste are estimated at approximately SEK 500 million (approximately US dollars 120 million) in terms of 1979 prices. The sum includes 25% contingency. The dismantling cost is equivalent to 10-15% of the installation cost of an equivalent new nuclear power plant. The exact percentage is dependent on the interest rate during the construction period. It is shown in the study that a total dismantling can be accomplished in less than five years. This report is a compilation of studies performed by ASEA-ATOM and VBB based on premises given by KBS. The reports from these studies are presented in appendices. (Auth.)

  16. Dismantling at the CEA's Nuclear Energy Division: strategy and programmes

    International Nuclear Information System (INIS)

    Lecomte, C.; Prunele, D. de; Rozain, J.P.; Nokhamzon, J.G.; Tallec, M.

    2008-01-01

    The CEA's Nuclear Energy Division (DEN) nuclear facilities currently include seventeen reactors and thirty six other miscellaneous facilities, particularly laboratories, fuel processing units and facilities specific to waste management. Some of these are currently being dismantled or must be dismantled soon so that the DEN, the Nuclear Energy Division, can construct new equipment and thus have available a range of R and D facilities in line with the issues of the nuclear industry of the future. At CEA, the first nuclear facility dismantling operations go back several dozen years and involve numerous and varied facilities. The first operations of any significance took place in the 1960's and 1970's and covered, for example, the first plutonium plant at Fontenay-aux-Roses (total dismantling) and small research reactors or critical models - CESAR and PEGGY at Cadarache and MINERVE at Fontenay-aux Roses (civil engineering cleaned up and kept). At La Hague, the dismantling of AT1, a pilot workshop used by the CEA during the 1970's to process irradiated fuels from fast neutron reactors, was completed in March 2001 (IAEA former stage 3, excluding civil engineering demolition). On the other hand, during this period of first dismantling, the intermediate-sized reactors (G1, Rapsodie) were only partially dismantled after shut down, mainly due to the lack of graphite and sodium waste management routes at the time. About twenty facilities were thus dealt with up to 2001, in other words about half of all the nuclear facilities shut down permanently before this date. (authors)

  17. A dual pressurized water reactor producing 2000 MWe

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  18. Dual shell pressure balanced reactor vessel. Final project report

    International Nuclear Information System (INIS)

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01

    The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

  19. Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

    International Nuclear Information System (INIS)

    Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

    1994-01-01

    This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

  20. Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

    International Nuclear Information System (INIS)

    Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

    1979-01-01

    A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

  1. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  2. Structural analysis of fuel rod applied to pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Danilo P.; Pinheiro, Andre Ricardo M.; Lotto, André A., E-mail: danilo.pinheiro@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The design of fuel assemblies applied to Pressurized Water Reactors (PWR) has several requirements and acceptance criteria that must be attended for licensing. In the case of PWR fuel rods, an important mechanical structural requirement is to keep the radial stability when submitted to the coolant external pressure. In the framework of the Accident Tolerant Fuel (ATF) program new materials have been studied to replace zirconium based alloys as cladding, including iron-based alloys. In this sense, efforts have been made to evaluate the behavior of these materials under PWR conditions. The present work aims to evaluate the collapse cold pressure of a stainless steel thin-walled tube similar to that used as cladding material of fuel rods by means of the comparison of numeric data, and experimental results. As a result of the simulations, it was observed that the collapse pressure has a value intermediate value between those found by regulatory requirements and analytical calculations. The experiment was carried out for the validation of the computational model using test specimens of thin-walled tubes considering empty tube. The test specimens were sealed at both ends by means of welding. They were subjected to a high pressure device until the collapse of the tubes. Preliminary results obtained from experiments with the empty test specimens indicate that the computational model can be validated for stainless steel cladding, considering the difference between collapse pressure indicated in the regulatory document and the actual limit pressure concerning to radial instability of tubes with the studied characteristics. (author)

  3. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  4. The Rossendorf research reactor. Operating and dismantling from a point of view of the emission control; Der Rossendorfer Forschungsreaktor. Betrieb und Rueckbau aus Sicht der Emissionsueberwachung

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, B.; Beutmann, A.; Kaden, M.; Scheibke, J. [VKTA, Dresden (Germany); Boessert, W.; Jansen, K.; Walter, M.

    2016-07-01

    The Rossendorf research reactor went in operation in 1957 as GDR's first nuclear reactor and Germanys second after FRM Garching. It was a heterogeneously structured, light-water moderated and cooled tank-reactor of the Soviet type WWR-S. During his time of operation, he served both the research and the production of radioisotopes. The history of exhaust air emission monitoring and its results are presented. With view to the decommissioning time selected results are discussed. The estimated discharges are compared by the actually recognized.

  5. Long-term management of wastes resulting from dismantling operations. Storing the very low-level activity wastes at Morvilliers

    International Nuclear Information System (INIS)

    Duret, F.; Dutzer, M.; Beranger, V.; Lecoq, P.

    2003-01-01

    Extension of dismantling operations in France in the years to come poses the question of availability of long-term waste facility. Large amount of such wastes will be produced after progressive shutdown of the 58 pressurized water reactors now in operation, not before 2010. However, France is already confronted with dismantling of 9 power reactors (6 of which of gas cooled graphite type), the first reprocessing plant at Marcoule, as well as, dismantling of other installations, for instance the CEA reactors or laboratories. The systems of processing the dismantling waste are not different from those used for wastes resulting from nuclear operations. For the high-level or long-term intermediate level activity disposal the debates must start by 2006, as based on the results of the research conducted according to different provisions of the December 30, 1991 law. These wastes represent however small amounts from the dismantling (around 2000 t for the 9 reactors at shutdown) and they will be stored until a decision will be made. A specific storing system should be implemented by 2008-2010 for the graphite wastes (around 23,000 t) which contain significant amount of long-lived radioelements, although their gross activity is low. But the most significant amount will come from low-level or intermediate-level of short lifetime or from wastes of very low activity. The first category is stored at Storage Center at Aube (CSA), its capacity being of 1,000,000 m 3 of drums. The total volume stored by the end of 2002 amounted 136,500 m 3 with an annual delivering of 12-15,000 m 3 at design rate of 30,000 m 3 /y. This center will be able to absorb the flux increase resulting from dismantling of the decommissioned nuclear installations (around 50,000 t from the dismantling of the 9 power reactor). The Center at Aube can be also adapted for storing wastes of large sizes as for instance the lid of the reactor vessel. According to the French regulation, the wastes produced within a

  6. Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

    International Nuclear Information System (INIS)

    Sun Yunxue; Shao Jin

    2015-01-01

    Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

  7. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  8. 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

  9. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Yung Joon

    1994-02-01

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  10. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  11. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1995-01-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

  12. In-place thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1985-04-01

    Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

  13. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  14. Dynamic loads on reactor vessel components by low pressure waves

    International Nuclear Information System (INIS)

    Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

    1978-01-01

    Starting from the conservation theorems for mass and impulses the code DRUWE has been developed enabling the calculation of dynamic loads of the reactor shell on the basis of simplified assumptions for the first period shortly after rupture. According to the RSK-guidelines it can be assumed that the whole weld size is opened within 15 msec. This time-dependent opening of the fractured plane can be taken into account in the computer program. The calculation is composed in a way that for a reactor shell devided into cross and angle sections the local, chronological pressure and strength curves, the total dynamic load as well as the moments acting on the fastenings of the reactor shell can be calculated. As input data only geometrical details concerning the concept of the pressure vessel and its components as well as the effective subcooling of the fluid are needed. By means of several parameters the program can be operated in a way that the results are available in form of listings or diagrams, respectively, but also as card pile for further examinations, e.g. strength analysis. (orig./RW) [de

  15. Thermal shield support degradation in pressurized water reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Fry, D.N.

    1986-01-01

    Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

  16. Current state of research on pressurized water reactor safety

    International Nuclear Information System (INIS)

    Couturier, Jean; Schwarz, Michel; Roubaud, Sebastien; Lavarenne, Caroline; Mattei, Jean-Marie; Rigollet, Laurence; Scotti, Oona; Clement, Christophe; Lancieri, Maria; Gelis, Celine; Jacquemain, Didier; Bentaib, Ahmed; Nahas, Georges; Tarallo, Francois; Guilhem, Gilbert; Cattiaux, Gerard; Durville, Benoit; Mun, Christian; Delaval, Christine; Sollier, Thierry; Stelmaszyk, Jean-Marc; Jeffroy, Francois; Dechy, Nicolas; Chanton, Olivier; Tasset, Daniel; Pichancourt, Isabelle; Barre, Francois; Bruna, Gianni; Evrard, Jean-Michel; Gonzalez, Richard; Loiseau, Olivier; Queniart, Daniel; Vola, Didier; Goue, Georges; Lefevre, Odile

    2018-03-01

    For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study - loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. -, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

  17. Consequence evaluation of hypothetical reactor pressure vessel support failure

    International Nuclear Information System (INIS)

    Lu, S.C.; Holman, G.S.; Lambert, H.E.

    1991-01-01

    This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

  18. Decontamination and dismantlement plan for international reviewing

    International Nuclear Information System (INIS)

    Wells, P.B.; Earle, O.K.; Klepikov, A.Kh.

    2000-01-01

    When developing a decommissioning plan, several factors need to be included. First and foremost is the issue of outline and scope. Specific to the BN-350, are issues related to short term tasks required to support the safe storage of the reactor for the next 50 years, and long term tasks required to dismantle the reactor, leaving some sort of final state, (brown field, green field, etc.) In addition, issues such as personnel and physical safety as well as environmental concerns must be addressed to ensure the shut down and dismantlement of the reactor is done in a safe manner, both for personnel and the environment. In addition to being the base document in which to support work, a D and D plan can also be utilized to obtain financial resources necessary to complete the plan, as is the case for the BN-350 Reactor located in Aktau, Kazakhstan. By providing a clear and complete D and D plan, which includes costs and schedules for each item, it is anticipated that donor countries will have the ability to review, approve, and provide financial support to complete the work described in the plan

  19. Overall strategy of Creys Malville power station dismantling

    International Nuclear Information System (INIS)

    Alphonse, P.

    2002-01-01

    The power station was stopped by a government decision following the elections in 1997. This shutdown was then made official by a letter dated April 1998 and the decree of December 1998. This was a non-technical shutdown and as such had not been envisaged; there has been no early warning. Current dismantling strategy: The studies leading to shutdown and then dismantling were engaged in 1998 based on a scenario with a status corresponding to IAEA level 2 until 2046. In 2001, EDF management made the decision to dismantle all the first generation power stations and Creys Malville between now and 2025. It should be noted that the presence of strongly irradiated stellite in the Creys Malville reactor would still require remote systems for working in the reactor block after 2046. The sequence of operations dictated by the dismantling strategy is as follows: eliminate the risks as soon as possible and in particular the risk related to the sodium, 3300 tonnes of which is kept in liquid form in the reactor vessel; dismantle the most active parts which are too radioactive to be sent to the existing or planned storage centres. This may lead to on-site storage to allow decay to occur before sending to a storage centre; planning of the work interventions in order to limit the costs

  20. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1988-04-01

    Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work

  1. LEP Dismantling: Wagons Roll!

    CERN Multimedia

    2001-01-01

    The first trucks transporting material from LEP and its four experiments left CERN on 31 January. Since the LEP dismantling operation began, the material had been waiting to be removed from the sites of the four experiments and the special transit area on the Prévessin site. On the evening of 30 January, the French customs authorities gave the green light for the transport operation to begin. So first thing the next day, the two companies in charge of recycling the material, Jaeger & Bosshard (Switzerland) and Excoffier (France), set to work. Only 1500 truckloads to go before everything has been removed!

  2. Clean-up and dismantling, Dismantling - legacy of the past, prospects for the future: CEA, a pioneer in the dismantling process, nuclear dismantling, research and innovation dedicated to dismantling

    International Nuclear Information System (INIS)

    Lorec, Amelie

    2016-01-01

    France - a world leader in the whole nuclear power cycle - is also responsible for the clean-up and dismantling of its end-of-life nuclear facilities. Here, the CEA is considered to be a pioneer both in the project ownership of work sites and in the R and D for optimising the timescales, costs and safety of those work sites. Its responsibilities range from defining the most appropriate scenario, characterising the radiological state of equipment and decontaminating premises, carrying out dismantling and optimising the resulting waste. With this wide range of skills and the diversity of its facilities, the CEA Nuclear Energy Division is developing innovative solutions which are already the subject of industrial transfers. Two-thirds of France's end-of-life nuclear facilities belong to the CEA - a situation connected with its history. This implies setting up clean-up and dismantling work sites which have unprecedented scientific, human and financial challenges. Every regulated nuclear installation (INB) (nuclear reactors, laboratories, etc.) has a limited operating life. When it stops being used, it is first cleaned up (removal of radioactive substances), then dismantled (disassembly of components) in accordance with the baseline safety requirements, and finally decommissioned so that it can be used for other purposes or be demolished. Cleanup and dismantling operations concern all the facility's components, such as hot (shielded) cells which can be found in some laboratories. As the owner of its clean-up and dismantling projects, the CEA also devotes a significant amount of R and D to reducing the timescales, costs and waste from current and future programmes, while improving their safety. The resulting innovations often lead to industrial transfers. (authors)

  3. Contemporary pressurized water reactor technology in the world

    International Nuclear Information System (INIS)

    Komarek, A.

    1991-01-01

    The recent political events enabled Czechoslovak industrial companies to come into direct contact with leading western companies involved in pressurized water ractor technology. A survey is presented of the present situation at the world market of PWR type nuclear power plant suppliers and suppliers of fuel cycle services. Information is given on the potential bids for the next Czechoslovak nuclear power plants with PWR reactors. Economic aspects of the potential bids are presented including some considerations about the participation of the Czechoslovak nuclear industry as a supplier of the reactor for the future power plants. Main technical parameters are listed of PWR units with an output about 1000 MW supplied by Westinghouse EC, ABB -Combustion Engineering and Siemens AG. At present, the bids for new Czechoslovak nuclear power plants are being evaluated. No information on terms of the bids actually coming from foreign companies is used in the article. (Z.S.). 9 figs., 5 tabs

  4. On-line fatigue monitoring system for reactor pressure vessel

    International Nuclear Information System (INIS)

    Tokunaga, K.; Sakai, A.; Aoki, T.; Ranganath, S.; Stevens, G.L.

    1994-01-01

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to an operating boiling water reactor (BWR), Tsuruga Unit-1, is described. The system uses the influence function approach and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computed fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification No.501. Fatigue usage results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension. (author)

  5. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  6. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  7. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    International Nuclear Information System (INIS)

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-01-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  8. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  9. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    International Nuclear Information System (INIS)

    Lin Chaung; Shen Chihming

    2000-01-01

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller

  10. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    GRIFFIN, PATRICK J.

    1999-01-01

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  11. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  12. Neutron irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Steele, L.E.

    1975-01-01

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  13. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  14. SURGTANK, Steam Pressure, Saturation Temperature or Reactor Surge Tank

    International Nuclear Information System (INIS)

    Gorman, D.J.; Gupta, R.K.

    2001-01-01

    1 - Description of problem or function: SURGTANK generates the steam pressure, saturation temperature, and ambient temperature history for a nuclear reactor steam surge tank (pressurizer) in a state of thermodynamic equilibrium subjected to a liquid insurge described by a specified time history of liquid levels. It is capable also of providing the pressure and saturation temperature history, starting from thermodynamic equilibrium conditions, for the same tank subjected to an out-surge described by a time history of liquid levels. Both operations are available for light- or heavy- water nuclear reactor systems. The tank is assumed to have perfect thermal insulation on its outer wall surfaces. 2 - Method of solution: Surge tank geometry and initial liquid level and saturation pressure are provided as input for the out-surge problem, along with the prescribed time-sequence level history. SURGTANK assumes a reduced pressure for the end of the first change in liquid level and determines the associated change of entropy for the closed system. The assumed pressure is adjusted and the associated change in entropy recalculated until a pressure is attained for which no change occurs. This pressure is recorded and used as the beginning pressure for the next level increment. The system is then re-defined to exclude the small amount of liquid which has left the tank, and a solution for the pressure at the end of the second level increment is obtained. The procedure is terminated when the pressure at the end of the final increment has been determined. Surge tank geometry, thermal conductivity, specific heat, and density of tank walls, initial liquid level, and saturation pressure are provided as input for the insurge problem, along with the prescribed time-sequence level history. SURGTANK assumes a slightly in- creased pressure for the end of the first level, the inner tank sur- face is assumed to follow saturation temperature, linearly with time, throughout the interval, and

  15. Applicability of AWJ technique for dismantling reactor of the Fukushima Daiichi Nuclear Power Station. Cutting test of imitation of fuel debris and optimization of the cutting condition

    International Nuclear Information System (INIS)

    Maruyama, Shin-ichiro; Watatani, Satoshi

    2016-01-01

    Based on findings during recovery works that followed the accident at Three Mile Island Station 2, it is assumed that the reactor internals at the Fukushima Daiichi Nuclear Power Station (1F) have complex geometries intermixed with melted fuel and confined in limited spaces. Accordingly, abrasive water jet (AWJ) cutting method is considered to be a promising technique that can be safely and reasonably used for cutting and removing reactor internals. The authors conducted tests to examine the possibility of application and to solve the problems of this technique. In the tests imitation of fuel debris and optimization of the cutting condition is used. The test result made the measures for some of the associated issues clear, and demonstrated that AWJ cutting method is assumed as one of the promising techniques for removing reactor internals. (author)

  16. Holographic NDE of pressure tubes for Cirene nuclear reactor

    International Nuclear Information System (INIS)

    Di Chirico, G.; Pirodda, L.; Villani, A.

    1985-01-01

    Pressure tubes for CIRENE nuclear reactor can be subjected to fretting corrosion of the inner walls. The resulting marks exhibit different geometries, whose influence on the structural behaviour of the tubes has been evaluated by means of a real time holographic technique. The paper shows the results of this investigation. Position and shape of internal defects have been directly visualized by observing holographic fringe distorsions on the outside surface of the tubes. Furthermore, through the fringe patterns, circumferential stress values have also been obtained. (Author) [pt

  17. Contribution to a neutronic calculation scheme for pressurized water reactors

    International Nuclear Information System (INIS)

    Martin Del Campo, C.

    1987-01-01

    This research thesis aims at developing and validating the set of data and codes which build up the neutron computation scheme of pressurized water reactors. More precisely, it focuses on the improvement of the precision of calculation of command clusters (absorbing components which can be inserted into the core to control the reactivity), and on the modelling of reflector representation (material placed around the core and reflecting back the escaping neutrons). For the first case, a precise calculation is performed, based on the transport theory. For the second case, diffusion constants obtained in the previous case and simplified equations are used to reduce the calculation cost

  18. Reactor pressure vessel embrittlement: Insights from neural network modelling

    Science.gov (United States)

    Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

    2018-04-01

    Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

  19. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Vasile, A.; Dufour, Ph.; Golfier, H.; Grouiller, J.P.; Guillet, J.L.; Poinot, Ch.; Youinou, G.; Zaetta, A.

    2003-01-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1 . More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  20. Automated ultrasonic shop inspection of reactor pressure vessel forgings

    International Nuclear Information System (INIS)

    Farley, J.M.; Dikstra, B.J.; Hanstock, D.J.; Pople, C.H.

    1986-01-01

    Automated ultrasonic shop inspection utilizing a computer-controlled system is being applied to each of the forgings for the reactor pressure vessel of the proposed Sizewell B PWR power station. Procedures which utilize a combination of high sensitivity shear wave pulse echo, 0 degrees and 70 degrees angled longitudinal waves, tandem and through-thickness arrays have been developed to provide comprehensive coverage and an overall reliability of inspection comparable to the best achieved in UKAEA defect detection trials and in PISC II. This paper describes the ultrasonic techniques, the automated system (its design, commissioning and testing), validation and the progress of the inspections

  1. Validation of the dynamic model for a pressurized water reactor

    International Nuclear Information System (INIS)

    Zwingelstein, Gilles.

    1979-01-01

    Dynamic model validation is a necessary procedure to assure that the developed empirical or physical models are satisfactorily representing the dynamic behavior of the actual plant during normal or abnormal transients. For small transients, physical models which represent isolated core, isolated steam generator and the overall pressurized water reactor are described. Using data collected during the step power changes that occured during the startup procedures, comparisons of experimental and actual transients are given at 30% and 100% of full power. The agreement between the transients derived from the model and those recorded on the plant indicates that the developed models are well suited for use for functional or control studies

  2. Environment sensitive cracking in light water reactor pressure boundary materials

    International Nuclear Information System (INIS)

    Haenninen, H.; Aho-Mantila, I.

    1985-01-01

    The purpose of the paper is to review the available methods and the most promising future possibilities of preventive maintenance to counteract the various forms of environment sensitive cracking of pressure boundary materials in light water reactors. Environment sensitive cracking is considered from the metallurgical, mechanical and environmental point of view. The main emphasis is on intergranular stress corrosion cracking of austenitic stainless steels and high strenght Ni-base alloys, as well as on corrosion fatigue of low alloy and stainless steels. Finally, some general ideas how to predict, reduce or eliminate environment sensitive cracking in service are presented

  3. Twin header bore welded steam generator for pressurized water reactors

    International Nuclear Information System (INIS)

    Davies, R.J.; Hirst, B.

    1979-01-01

    A description is given of a pressurized water reactor (PWR) steam generator concept, several examples of which have been in service for up to fourteen years. Details are given of the highly successful service record of this equipment and the features which have been incorporated to minimize corrosion and deposition pockets. The design employs a vertical U tube bundle carried off two horizontal headers to which the tubes are welded by the Foster Wheeler Power Products (FWPP) bore welding process. The factors to be considered in uprating the design to meet the current operating conditions for a 1000 MW unit are discussed. (author)

  4. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

  5. Improved plenum pressure gradient facemaps for PKL reactors

    International Nuclear Information System (INIS)

    Crowley, D.A.; Hamm, L.L.

    1988-05-01

    This report documents the development of improved plenum pressure gradient facemaps* for PKL Mark 16--31 and Mark 22 reactor charges. These new maps are based on the 1985 L-area AC flow tests. Use of the L-area data base for estimating C-area plenum pressure gradient maps is inappropriate because the nozzle geometry plays a major role in determining the shape of the plenum pressure profile. These plenum pressure gradient facemaps are used in the emergency cooling system (ECS) and in the flow instability (FI) loss of coolant accident (LOCA) limits calculations. For the ECS LOCA limits calculations, the maps are used as input to the FLOWZONE computer code to determine the average flow within a flowzone during normal operating conditions. For the FI LOCA limits calculations, the maps are used as plenum pressure boundary conditions in the FLOWTRAN computer code to determine the maximum pre-incident assembly flow within a flowzone. These maps will also be used for flowzoning and transient protection limits analyses

  6. Dismantling of transuranic contaminated facilities

    International Nuclear Information System (INIS)

    Roux, P.

    1985-01-01

    The dismantling of transuranic contaminated facilities raises specific problems. A large part of these problems relates to the management of the waste resulting from dismantling. From the experience gained in the different centers CEA and COGEMA it appears that there are industrial solutions in the group CEA and that an engineering company such as SGN can export them [fr

  7. Machine for dismantling metal parts

    Energy Technology Data Exchange (ETDEWEB)

    Prokopov, O.I.; Loginovskiy, V.I.; Yagudin, S.Z.

    1982-01-01

    The purpose of the invention is to reduce the outlays of time for dismantling metal parts under conditions of eliminating open gas and oil gushers in operational drilling. This goal is achieved because the machine for dismantling the metal parts is equipped with a set of clamping elements arranged on the chassis, where each of them has a drive.

  8. 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.)

  9. Robotics take heat out of reactor. [Windscale AGR decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Rufford, N

    1986-12-04

    The Windscale prototype reactor is being decommissioned and dismantled. The stages are outlined. The first phase began in 1985 and included construction of a waste packaging plant annexed to the steel dome. The boilers will be cut up and, once decontaminated, probably sold for scrap. The second phase involves dismantling the reactor itself. Much of this will be done by a semi-automatic robot which is being specially built and tested. The robot will have an extendable arm with a manipulator which will be equipped with bolt croppers, shears, a saw and oxypropane cutter. This robot will cut up the pressure vessel in sections ready for encasing in concrete. Lessons learnt from the dismantling will be used in future reactor designs and specifications (eg the need to use steels with fewer impurities, especially cobalt). Ultimate disposal of the concrete waste blocks is undecided. (U.K.).

  10. Dismantling and waste management: CEA's strategy and research programs

    International Nuclear Information System (INIS)

    Behar, C.

    2012-01-01

    There are 3 main dismantling operations in CEA. First, the dismantling of the UP1 facility in the Marcoule site. UP1 was a reprocessing plant of nuclear fuels that operated from 1958 to 1997 and is now the biggest dismantling operation in the world. Its dismantling operation follows a 6-step scheme that will end in 2050. Secondly, the Passage project on the Grenoble site that concerns the dismantling of 3 research reactors (Siloette, Melusine and Siloe), of a laboratory dedicated to the analysis of active materials (Lama) and of a station for the processing of waste (Sted). Thirdly the Aladin project that concerns the installations of the Fontenay-aux-Roses site. The dismantling operations are complex because all the first research programs on high activity chemistry and on transuranium elements were performed in Fontenay-aux-Roses facilities and because ancient activities have to leave a clean place to be replaced by new ones. The radioactive waste produced by CEA enter the flow of waste that is normally processed and managed by ANDRA. Only high-activities waste have not yet a definitive solution, they are stored in waiting the opening of a geological repository. CEA leads research programs on the separation and transmutation of minor actinides and on the long-term behaviour of waste packages put in deep geological layers. (A.C.)

  11. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  12. Reliability aspects of radiation damage in reactor pressure vessel mterials

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1985-01-01

    The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

  13. Repairing method for shroud in reactor pressure vessel

    International Nuclear Information System (INIS)

    Watanabe, Yusuke.

    1996-01-01

    The present invention provides a method of repairing a shroud disposed in a pressure vessel of a BWR type reactor. Namely, a baffle plate is disposed on the outer surface of the lower portion of the shroud supported by a shroud support of the pressure vessel. The baffle plate is connected with a lug for securing a shroud head bolt disposed on the outer surface of an upper portion of the shroud by reinforcing members. With such a constitution, when crackings are caused in the shroud, the development of the crackings can be prevented without losing the function of securing the shroud head bolt. Further, if a material having thermal expansion coefficient lower than that of austenite stainless steel is used for the material of the reinforcing member, clamping load to be applied upon attaching the auxiliary member can be reduced. As a result, operation for the attachment is facilitated. (I.S.)

  14. Probabilistic structural integrity of reactor vessel under pressurized thermal shock

    International Nuclear Information System (INIS)

    Myung Jo Hhung; Young Hwan Choi; Hho Jung Kim; Changheui Jang

    2005-01-01

    Performed here is a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel. A round robin consisting of 1 prerequisite study and 5 cases for probabilistic approaches is proposed, and all organizations interested are invited. The problems are solved and their results are compared to issue some recommendation of best practices in this area and to assure an understanding of the key parameters of this type of approach, which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria. Six participants from 3 organizations in Korea responded to the problem and their results are compiled in this study. (authors)

  15. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  16. Development of project management data calculation models relating to dismantling of nuclear facilities. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Sukegawa, Takenori; Ohshima, Soichiro; Shiraishi, Kunio; Yanagihara, Satoshi [Department of Decommissioning and Waste Management, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai Ibaraki (Japan)

    1999-02-01

    Labor-hours necessary for dismantling activities are generally estimated based on experience, for example, as a form of unit productivity factors such as the relationship between labor-hours and weight of components dismantled which were obtained by actual dismantling activities. The project management data calculation models together with unit productivity factors for basic dismantling work activities were developed by analyzing the data obtained from the Japan Power Demonstration Reactor (JPDR) dismantling project, which will be applicable to estimation of labor-hours in various dismantling conditions. Typical work breakdown structures were also prepared by categorizing repeatable basic dismantling work activities for effective planning of dismantling activities. The labor-hours for dismantling the JPDR components and structures were calculated by using the code system for management of reactor decommissioning (COSMARD), in which the work breakdown structures and the calculation models were contained. It was confirmed that the labor-hours could be easily estimated by COSMARD through the calculations. This report describes the labor-hour calculation models and application of these models to COSMARD. (author)

  17. Development of decommissioning engineering support system for fugen. Development of support system during actual dismantlement works

    International Nuclear Information System (INIS)

    Masanori Izumi; Yukihiro Iguchi; Yoshiki Kannehira

    2005-01-01

    The Advanced Thermal Reactor, Fugen Nuclear Power Station was permanently shut down in March 2003, and is now preparing for decommissioning. We have been developing Decommissioning Engineering Support System (DEXUS) aimed at planning optimal dismantlement process and carrying out dismantlement work safely and precisely. DEXUS consists of 'decommissioning planning support system' and 'dismantling support system'. The dismantling support system is developed aiming at using during actual dismantling work. It consists of three subsystems such as 'Worksite Visualization System', 'Dismantling Data Collection System' and 'Generated Waste Management System'. 'Worksite Visualization System' is a support system designed to provide the necessary information to workers during actual dismantlement works. And this system adopts AR (Augmented Reality) technology, overlapping calculation information into real world. 'Dismantling Data Collection System' is to collect necessary data for improving accuracy of decommissioning planning by evaluating work content and worker equipage, work time for dismantlement works. 'Generated Waste Management system' is a system recording necessary information by attaching the barcode to dismantled wastes or the containers. We can get the information of generated waste by recording generation place, generated time, treatment method and the contents. These subsystems enable to carry out reasonable and safe decommissioning of Fugen. In addition, we expect that those systems will be used for decommissioning of other nuclear facilities in the future. (authors)

  18. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-01-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  19. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  20. Fuzzy power control algorithm for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Y.J.; Lee, B.W.

    1994-01-01

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

  1. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-01

    The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

  2. Design and safety of the Sizewell pressurized water reactor

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the UK the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear power focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The long-term consequences can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking. (author)

  3. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  4. Inspection and repair of reactor pressure vessel (RPV) internals

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.; Nicolai, M.

    1996-01-01

    The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

  5. Biofilm architecture in a novel pressurized biofilm reactor.

    Science.gov (United States)

    Jiang, Wei; Xia, Siqing; Duan, Liang; Hermanowicz, Slawomir W

    2015-01-01

    A novel pure-oxygen pressurized biofilm reactor was operated at different organic loading, mechanical shear and hydrodynamic conditions to understand the relationships between biofilm architecture and its operation. The ultimate goal was to improve the performance of the biofilm reactor. The biofilm was labeled with seven stains and observed with confocal laser scanning microscopy. Unusual biofilm architecture of a ribbon embedded between two surfaces with very few points of attachment was observed. As organic loading increased, the biofilm morphology changed from a moderately rough layer into a locally smoother biomass with significant bulging protuberances, although the chemical oxygen demand (COD) removal efficiency remained unchanged at about 75%. At higher organic loadings, biofilms contained a larger fraction of active cells distributed uniformly within a proteinaceous matrix with decreasing polysaccharide content. Higher hydrodynamic shear in combination with high organic loading resulted in the collapse of biofilm structure and a substantial decrease in reactor performance (a COD removal of 16%). Moreover, the important role of proteins for the spatial distribution of active cells was demonstrated quantitatively.

  6. Behavior of stainless steels in pressurized water reactor primary circuits

    International Nuclear Information System (INIS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-01-01

    Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

  7. Self-pressurization analysis of the natural circulation integral nuclear reactor using a new dynamic model

    Directory of Open Access Journals (Sweden)

    Ali Farsoon Pilehvar

    2018-06-01

    Full Text Available Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established. Keywords: Condensation Power, Flashing Phenomenon, Natural Circulation, Self-Pressurization, Small Modular Reactor

  8. Remote controlled stud bolt handling device for reactor pressure vessel

    International Nuclear Information System (INIS)

    Shindo, Takenori; Shigehiro, Katsuya; Ito, Morio; Okada, Kenji

    1988-01-01

    In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

  9. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  10. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  11. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  12. Improvement of Algorithms for Pressure Maintenance Systems in Drum-Separators of RBMK-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aleksakov, A. N., E-mail: yankovskiy.k@nikiet.ru; Yankovskiy, K. I. [JSC “N. A. Dollezhal Research and Development Institute of Power Engineering (NIKIET),” (Russian Federation); Dunaev, V. I.; Kushbasov, A. N. [JSC “Diakont,” (Russian Federation)

    2015-05-15

    The main tasks and challenges for pressure regulation in the drum-separators of RBMK-1000 reactors are described. New approaches to constructing algorithms for pressure control in drum-separators by electro-hydraulic turbine control systems are discussed. Results are provided from tests of the operation of modernized pressure regulators during fast transients with reductions in reactor power.

  13. Study of the origin of elements of the uranium-235 family observed in excess in the vicinity of the experimental nuclear EL4 reactor under dismantling. Lessons got at this day and conclusions

    International Nuclear Information System (INIS)

    2007-01-01

    This study resumes the discovery of an excess of actinium 227 found around by EL4 nuclear reactor actually in dismantling. The search for the origin of this excess revealed a real inquiry of investigation during three years. Because a nuclear reactor existed in this area a particular attention will have concerned this region. The doubt became the line of conduct to find the answer to the human or natural origin of this excess. Finally and against any evidence, it appears that the origin of this phenomenon was natural, consequence of the particular local geology. The detail of the different investigations is given: search of a possible correlation with the composition of elevations constituent of lanes, search (and underlining) of new sites in the surroundings of the Rusquec pond and the Plouenez station, study of the atmospheric deposits under winds of the nuclear power plant and in the east direction, search of a possible relationship with the gaseous effluents of the nuclear power plant in the past, historical study of radioactive effluents releases in the fifty last years by the analysis of the sedimentary deposits in the Saint-Herbiot reservoir, search of a possible correlation between the excess of actinium 227 and the nuclear power plant activity; search of a possible correlation with a human activity without any relationship with the nuclear activities, search of a correlation with the underground waters, search of a correlation with the geological context, collect of information on the possible transfers in direction of the food chain, determination of the radiological composition of the underground waters ( not perturbed by human activity), search of the cause of an excess of actinium 227 in the old channel of liquid effluents release of the nuclear power plant. The results are given and discussed. And contrary to all expectations the origin of the excess of actinium 227 is completely natural. (N.C.)

  14. Optimisation of radiation protection for the new european pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Miniere, D.; Beneteau, Y.; Le Guen, Bernard

    2008-01-01

    Full text: As part of the EPR (European Pressurized Reactor) project being deployed at Flamanville, EDF has pro actively made the decision to focus on radiation protection (RP) aspects right from the start of the design phase, as it has done with nuclear safety. The approach adopted for managing RP-significant activities has been to include all involved stake holders -designers, licensee and contractor companies- in the three successive phases, starting with a survey among workers and designers, followed by a proposal review, and finally ending with the decision-making phase entrusted to an ALARA committee. The RP target set by EDF for this new reactor is to engage in an effort of continuous improvement and optimisation, through benchmarking with the best performing plants of the fleet. The collective dose target is currently set at 0.35 man.Sv/year per unit. In addition to other aspects, efforts will focus on shortening the duration of the highest-dose jobs, with a new challenge being set for work performed in the reactor building during normal operations, the aim being to improve plant availability. The plan is for work to be performed 7 days prior to shutting down the reactor and 3 days afterwards, in order to make logistical arrangements for forthcoming jobs. Without this reduction, the estimated drop is currently 4.5% of annual dose. For this purpose, two areas have been set up in the EPR 's reactor building: one no-go area for containing leaks from the primary circuit, and one accessible area for normal operations, separated from the no-go area by purpose-built ventilation equipment and facilities. To offer protection against radioactive flux (neutrons and high energy), RP studies have resulted in the installation of a concrete floor and of nuclear shielding at the outlets of primary circuit pipes. Steam generator bunkers and pumps have also been reinforced. All these measures will ensure that the accessible area can be posted as a green area (dose rate < 25

  15. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1987-01-01

    This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals and gas bubble nucleation in molten metals are relevant problems which are addressed in this work. Models are developed for jet expansion, primary breakup of the jet and secondary fragmentation of melt droplets resulting from violent effervescence of dissolved gas. The jet expansion model is based on a general relation for bubble growth which includes both inertia-controlled and diffusion-controlled growth phases. The jet expansion model is able to predict the jet void fraction, jet radius as a function of axial distance from the pressure vessel, bubble size and bubble pressure. The number density of gas bubbles in the melt, which is a basic parameter in the model, was determined experimentally and is about 10 8 per m 3 of liquid. The primary breakup of the jet produces a spray of droplets, about 2-3 mm in diameter. Parametric calculations for a TMLB' reactor accident sequence show that the corium jet is disrupted within a few initial jet diameters from the reactor vessel and that the radius of corium spray at the level of the reactor cavity floor is in the range of 0.8 to 2.6 m. (orig./HP)

  16. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    Science.gov (United States)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  17. Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

    For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

  18. Expertise of the Oeko-Institute on the application to obtain permission to partially dismantle the Niederaichbach nuclear power plant

    International Nuclear Information System (INIS)

    1985-01-01

    This expertise gives an overview on the problems associated with the decommissioning and dismantling of the Niederaichbach nuclear power plant, considering technical and legal aspects. It wants to prove that the dismantling of this reactor cannot serve as evidence to prove the general feasibility of reactor dismantling. Much space is dedicated to the discussion about where the borderline should be drawn between radioactive and non-radioactive materials according to the ordinance on radiation protection. The reasons for rejecting the partial dismantling application are given. (DG) [de

  19. Decommissioning of a small reactor (BR3 reactor, Belgium)

    International Nuclear Information System (INIS)

    Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

    2002-01-01

    Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

  20. Development of design technology for an advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Kim, Dong Soo; Chang, Won Pyo; Park, Koon Chul

    1991-07-01

    The objective of the project is to localize technology for the improvement of the reactor coolant system through a multidimensional thermal-hydraulic analysis for the steam generator and the pressurizer. Flow distribution analysis has been done for the YGN 3/4 steam generators when steady-state output conditions were varied in the ranges such as 100, 75, 50, and 25 using three-dimensional ATHOS 3 code. The results of the thermal-hydraulic analysis have been used for flow-induced vibration analysis for the YGN 3/4 steam generators. ATHOS 3 code has been modified for YGN 3/4 steam generator tube lane region using the cartesian geometry and the local porosity in the boundaries of the two adjacent cells. Stability ratio for the tube vibration has been calculated the modified ATHOS 3 and ANSYS code. A sensitivity study for the pressurizer volume change has been analyzed using LTC code which is for the performance analysis to predict an optimistic pressurizer volume. (Author)

  1. East/west steels for reactor pressure vessels

    International Nuclear Information System (INIS)

    Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

    1997-01-01

    The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

  2. Design study on steam generator integration into the VVER reactor pressure vessel

    International Nuclear Information System (INIS)

    Hort, J.; Matal, O.

    2004-01-01

    The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

  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. Materials characterization for advanced pressurized water reactors: Pt. 2

    International Nuclear Information System (INIS)

    Little, E.A.; Gage, G.

    1994-01-01

    A compilation and overview is presented of the experimental techniques available for characterization of the microstructural changes induced by neutron irradiation of PWR pressure vessel steels, and directed towards monitoring of embrittlement processes by examination of surveillance samples from advanced reactor systems. The microstructural features of significance include copper precipitation, dislocation loop and/or microvoid matrix damage and grain boundary solute segregation. The techniques of transmission electron microscopy, field-emission gun scanning transmission electron microscopy, small angle neutron scattering, positron annihilation and field-ion microscopy have all developed to a degree of sophistication such that they are capable of providing detailed microstructural information in these areas, and afford considerable insight into embrittlement processes when used in combination. (author)

  5. Reactor Pressure Vessel P-T Limit Curve Round Robin

    Energy Technology Data Exchange (ETDEWEB)

    Jang, C.H.; Moon, H.R.; Jeong, I.S. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    This report is the summary of the analysis results for the P-T Limit Curve construction which have been subjected to the round robin analysis. The purpose of the round robin is to compare the procedure and method used in various organizations to construct P-T limit curve to prevent brittle fracture of reactor pressure vessel of nuclear power plants. Each Participant used its own approach to construct the P-T limit curve and submitted the results, By analyzing the results, the reference procedure for the P-T limit curve could be established. This report include the results of the comparison of the procedure and method used by the participants, and sensitivity study of the key parameters. (author) 23 refs, 88 figs, 17 tabs.

  6. Pressurized water reactor iodine spiking behavior under power transient conditions

    International Nuclear Information System (INIS)

    Ho, J.C.

    1992-01-01

    The most accepted theory explaining the cause of pressurized water reactor iodine spiking is steam formation and condensation in damaged fuel rods. The phase transformation of the primary coolant from water to steam and back again is believed to cause the iodine spiking phenomenon. But due to the complex nature of the phenomenon, a comprehensive model of the behavior has not yet been successfully developed. This paper presents a new model based on an empirical approach, which gives a first-order estimation of the peak iodine spiking magnitude. Based on the proposed iodine spiking model, it is apparent that it is feasible to derive a correlation using the plant operating data base to monitor and control the peak iodine spiking magnitude

  7. Computer code for simulating pressurized water reactor core

    International Nuclear Information System (INIS)

    Serrano, A.M.B.

    1978-01-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

  8. Code for the core simulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1978-08-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt

  9. Simulation of pressurized water reactor in accidental state

    International Nuclear Information System (INIS)

    Chakir, E.

    1994-01-01

    The aim of this work is to develop the 1300 MWe 4 loops 'PWR' simulator called 'SATRAPE', witch the adopted physics modelisation allows a simplified neutronic calculation, and focus essentially on the reactor thermal hydraulic behavior in the case of the following accidents: - Loss of Coolant Accident (LOCA). - Steam Generator Tube Failure (SGTF). - Steam Line Break (SLB). In case of the 'LOCA' or 'SLB' accident, this modelisation enables the calculation of the pressure and the temperature in the containment building, and also the debit of the released dose in this latter in case of the 'LOCA' accident. The adopted models are relatively simple so as to allow an explicit resolve. In SATRAPE, two graphical interfaces enables to launch orders, whereas the other permits to visualize, the principal state variables of installations. The results obtained show a very good consistency with the envisaged commonly scenario at the time of the considered accidents. 33 refs., 52 figs., 1 tab. (author)

  10. Diagnostic system for primary circuits of pressurized-water reactors

    International Nuclear Information System (INIS)

    Liska, J.; Majer, J.

    1983-01-01

    The diagnostic system monitors the reactor, the main circulating pipe, the main circulating pump, the main shut-off valve, the steam generator and the pressurizer. Diagnostic signals are obtained from the sensors designed for operation measurements and from sensors for special diagnostic purposes. The following operations are carried out: detection of dangerous dynamic stress of components, detection of damage to functional surfaces of components, detection of occurrence and propagation of defects in component materials, detection of loose particles and foreign bodies, detection of coolant leakage, detection of coolant boiling in the core and detection of impermissible non-homogeneities of fields of physical quantities in the core. The diagnostic system comprises: monitoring, classification of properly investigated effects, periodical tracing and long-term tracing. The operational diagnostics system developed by the SKODA Concern consists of a vibration monitoring system, a spectral analysis system and a central evaluation system. (M.D.)

  11. Reactor pressure elevation preventing device upon interruption of load

    International Nuclear Information System (INIS)

    Ota, Yasuo; Okukawa, Ryutaro.

    1996-01-01

    In a power load imbalance circuit of a steam turbine control device, a power load imbalance occurrence signal is outputted for a predetermined period of time upon occurrence of load interruption. A function for suppressing increase of number of rotation of a turbine due to load interruption is not disturbed, and the power load imbalance circuit is not operated at least after a primary peak where the number of rotation of the turbine is increased. Since a steam control valve flow rate demand signal and a turbine bypass valve flow rate demand signals are corporated subsequently to control the opening degree of the steam control valve and the turbine bypass valve, elevation of reactor pressure is always suppressed and maintained constant, as well as abrupt opening of the steam control valve due to cancel of the power load imbalance circuit when steam control valve opening demand is outputted can be prevented. (N.H.)

  12. Ductile fracture estimation of reactor pressure vessel under thermal shock

    International Nuclear Information System (INIS)

    Takahashi, Jun; Sakai, Shinsuke; Okamura, Hiroyuki

    1990-01-01

    This paper presents a new scheme for the estimation of unstable ductile fracture of a reactor pressure vessel under thermal shock conditions. First, it is shown that the bending moment applied to the cracked section can be evaluated by considering the plastic deformation of the cracked section and the thermal deformation of the shell. As the contribution of the local thermal stress to the J-value is negligible, the J-value under thermal shock can be easily evaluated by using fully plastic solutions for the cracked part. Next, the phenomena of ductile fracture under thermal shock are expressed on the load-versus-displacement diagram which enables us to grasp the transient phenomena visually. In addition, several parametrical surveys are performed on the above diagram concerning the variation of (1) thermal shock conditions, (2) initial crack length, and (3) J-resistance curve (i.e. embrittlement by neutron irradiation). (author)

  13. Aging assessment of PWR [Pressurized Water Reactor] Auxiliary Feedwater Systems

    International Nuclear Information System (INIS)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab

  14. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  15. Design and analysis of pressurized water reactor systems

    International Nuclear Information System (INIS)

    Juhn, P.E.; Kim, Y.H.

    1979-01-01

    To help develop nuclear engineering technologies in local industry sectors, technical and economical data on pressurized water reactor systems and components have been collected, systematically analyzed and computerized to a certain degree. Codes and standards necessary for engineering design of PWR systems have been surveyed and clarified in terms of NSSS, turbine-generator system and BOP, then again rearranged with respect to quality classes and seismic classes. Some design manuals, criteria and guidelines regarding design, construction, test and operation of PWR plants have also been surveyed and collected. Benchmark cost calculation for the construction of a 900 MWe PWR plant, according to the standard format, was carried out, and computer model on construction costs was improved and updated by considering the local supply of labor and materials. And for the indigeneous development of PWR equipment and materials, such data as delivery schedule and manufacturers of 52 systems and 36,000 components have also been reviewed herein. (author)

  16. Preparation of the Shippingport reactor pressure vessel shipping package

    International Nuclear Information System (INIS)

    Yannitell, D.M.

    1988-01-01

    Shippingport Station Decommissioning Project is the removal and shipment the Reactor Pressure Vessel (PRV) and its associated Neutron Shield Tank (NST) to the government owned Hanford Reservation in Richland, Washington. Engineering studies considered the alternatives for removal and shipment of the RPV/NST. These included segmentation for subsequent truck shipments, and one-piece removal with barge or rail shipment. Although the analysis indicated that current technology could be utilized to accomplish either alternative, one-piece removal of the RPV was selected as the safest, most cost effective method. When compared to segmentation, it was estimated that one-piece removal would reduce the duration of the Project by 1 year, reduce cost by $4 M, and result in a savings of radiation exposure of 150 man-Rem. Rail transportation of an integral RPV/NST package is not feasible due to the physical size of the package. 5 refs., 1 fig

  17. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  18. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Berg, S.; Loeseth, S.; Holand, I.

    1977-01-01

    A computational model for circular symmetric reinforced concrete shell problems is described. The model is based on the Finite Element Method. Non-linear stress-strain constitutive relations are used for the concrete, the reinforcement and for the liner. The reinforcement layers may be of different steel qualities. Each layer may be given a specified prestressing. This can be done at the beginning of the computations or the specific reinforcement layer can be considered inactive until a specified level of loading is reached. Thus, the prestressing procedure may also be analyzed in detail. Bond-slip effects are not accounted for. However, no bond may be assumed for prestressing cables by inserting special reinforcement elements. Several models of prestressed concrete reactor pressure vessels which have been tested up to rupture have been analysed. Analytical (numerical) models for reinforced concrete are also discussed on a more general basis. (Auth.)

  19. Computer system for International Reactor Pressure Vessel Materials Database support

    International Nuclear Information System (INIS)

    Arutyunjan, R.; Kabalevsky, S.; Kiselev, V.; Serov, A.

    1997-01-01

    This report presents description of the computer tools for support of International Reactor Pressure Vessel Materials Database developed at IAEA. Work was focused on raw, qualified, processed materials data, search, retrieval, analysis, presentation and export possibilities of data. Developed software has the following main functions: provides software tools for querying and search of any type of data in the database; provides the capability to update the existing information in the database; provides the capability to present and print selected data; provides the possibility of export on yearly basis the run-time IRPVMDB with raw, qualified and processed materials data to Database members; provides the capability to export any selected sets of raw, qualified, processed materials data

  20. 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

  1. Evolution of general design requirements for french pressurized water reactors

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Rollinger, F.

    1988-10-01

    The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5

  2. Study on Material Selection of Reactor Pressure Vessel of SCWR

    Science.gov (United States)

    Ma, Shuli; Luo, Ying; Yin, Qinwei; Li, Changxiang; Xie, Guofu

    This paper first analyzes the feasibility of SA-508 Grade 3 Class 1 Steel as an alternative material for Supercritical Water-Cooled Reactor (SCWR) Reactor Pressure Vessel (RPV). This kind of steel is limited to be applied in SCWR RPV due to its quenching property, though large forging could be accomplished by domestic manufacturers in forging aspect. Therefore, steels with higher strength and better quenching property are needed for SWCR RPV. The chemical component of SA-508 Gr.3 Cl.2 steel is similar to that of SA-508 Gr.3 Cl.1 steel, and more appropriate matching of strength and toughness could be achieved by the adjusting the elements contents, as well as proper control of tempering temperature and time. In light of the fact that Cl.2 steel has been successfully applied to steam generator, it could be an alternative material for SWCR RPV. SA-508 Gr.4N steel with high strength and good toughness is another alternative material for SCWR RPV. But large amount of research work before application is still needed for the lack of data on welding and irradiation etc.

  3. Taipower's reload safety evaluation methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Ping-Hue; Yang, Y.S.

    1996-01-01

    For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

  4. Technical challenges for dismantlement verification

    International Nuclear Information System (INIS)

    Olinger, C.T.; Stanbro, W.D.; Johnston, R.G.; Nakhleh, C.W.; Dreicer, J.S.

    1997-01-01

    In preparation for future nuclear arms reduction treaties, including any potential successor treaties to START I and II, the authors have been examining possible methods for bilateral warhead dismantlement verification. Warhead dismantlement verification raises significant challenges in the political, legal, and technical arenas. This discussion will focus on the technical issues raised by warhead arms controls. Technical complications arise from several sources. These will be discussed under the headings of warhead authentication, chain-of-custody, dismantlement verification, non-nuclear component tracking, component monitoring, and irreversibility. The authors will discuss possible technical options to address these challenges as applied to a generic dismantlement and disposition process, in the process identifying limitations and vulnerabilities. They expect that these considerations will play a large role in any future arms reduction effort and, therefore, should be addressed in a timely fashion

  5. Investigation of the failure of a reactor pressure vessel by plastic instability

    International Nuclear Information System (INIS)

    Laemmer, H.; Ritter, B.

    1994-01-01

    A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

  6. Definition of a dismantling project

    International Nuclear Information System (INIS)

    Meyers, H.; Claes, J.; Geens, L.

    1988-01-01

    The shutdown of the fuel reprocessing plant of Eurochemic having been decided, a study for defining the facilities to be dismantled and how to do it, was conducted by Belgoprocess. The cost of the operation was estimated by an accurate investigation and by a pilot project on the dismantling of the wastes storage building. The work carried out up to now and the problems to be solved are summarized [fr

  7. BRET fuel assembly dismantling machine

    International Nuclear Information System (INIS)

    Titzler, P.A.; Bennett, K.L.; Kelley, R.S. Jr.; Stringer, J.L.

    1984-08-01

    An automated remote nuclear fuel assembly milling and dismantling machine has been designed, developed, and demonstrated at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The machine can be used to dismantle irradiated breeder fuel assemblies from the Fast Flux Test Facility prior to fuel reprocessing. It can be installed in an existing remotely operated shielded hot cell facility, the Fuels and Materials Examination Facility (FMEF), at the Hanford Site in Richland, Washington

  8. Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

    International Nuclear Information System (INIS)

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  9. Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jooho, W.; Baldwin, G. T.

    2005-04-01

    One critical aspect of any denuclearization of the Democratic People’s Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for “complete, verifiable and irreversible dismantlement,” or “CVID.” It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long

  10. The Grenoble CEA Center: dismantled and rehabilitated

    International Nuclear Information System (INIS)

    Anon.

    2013-01-01

    The denuclearization program of the CEA center in Grenoble was launched in 2001. It involves 6 nuclear facilities (3 research reactors: Melusine, Siloette, and Siloe, and 1 laboratory (LAMA) and 2 units for processing wastes). The dismantling works were finished at the end of 2012 and the 2013 program concerns: the demolition of the buildings homing Melusine and Siloe reactors, the final rehabilitation of the Siloe raft, and the final rehabilitation of the laboratory and of the waste processing units. The budget is 117*10 6 euros for Siloe, 28*10 6 euros for Melusine, 6*10 6 euros for Siloette, 70*10 6 euros for the LAMA, and 90*10 6 euros for the 2 waste processing units. (A.C.)

  11. Evaluation of worker's dose on a virtual dismantling environment

    International Nuclear Information System (INIS)

    Park, Hee Seong; Kim, Sung Hyun; Park, Byung Suk; Yoon, Ji Sup

    2007-01-01

    The motivation of this study is to provide a basis for a minimization of worker's dose during dismantling activities. In the present study, we proposed methods for identifying an existence of radioactivity which is contained in the dismantling objects and for evaluating a worker's dose under a virtual dismantling environment. To evaluate a worker's external dose, the shape of the exposure room in the KRR 2(Korean Research Reactor TRIGA MARK III) by 3D CAD was created and the radiation dose surrounding the facility by using MCNP- 4C(Monte Carlo N-Particle-4C) was calculated. The radiation field of the exposure room was visualized three dimensionally by using the radiation dose that was obtained by the code

  12. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Connor, J.J.

    1975-01-01

    The numerical procedures for predicting the nonlinear behavior of a prestressed concrete reactor vessel over its design life are discussed. The numerical models are constructed by combining three-dimensional isoparametric finite elements which simulate the concrete, thin shell elements which simulate steel linear plates, and layers of reinforcement steel, and axial elements for discrete prestressing cables. Nonlinearity under compressive stress, multi-dimensional cracking, shrinkage and stress/temperature induced creep of concrete are considered in addition to the elasti-plastic behavior of the liner and reinforcing steel. Various failure theories for concrete have been proposed recently. Also, there are alternative strategies for solving the discrete system equations over the design life, accounting for test loads, pressure and temperature operational loads, creep unloading and abnormal loads. The proposed methods are reviewed, and a new formulation developed by the authors is described. A number of comparisons with experimental tests results and other numerical schemes are presented. These examples demonstrate the validity of the formulation and also provide valuable information concerning the cost and accuracy of the various solution strategies i.e., total vs. incremental loading and initial vs. tangent stiffness. Finally, the analysis of an actual PCRV is described. Stress contours and cracking patterns in the region of cutouts corresponding to operational pressure and temperature loads are illustrated. The effects of creep, unloading, and creep recovery are then shown. Lastly, a strategy for assessing the performance over its design life is discussed

  13. Minimum weight design of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Boes, R.

    1975-01-01

    A method of non-linear programming for the minimization of the volume of rotationally symmetric prestressed concrete reactor pressure vessels is presented. It is assumed that the inner shape, the loads and the degree of prestressing are prescribed, whereas the outer shape is to be detemined. Prestressing includes rotational and vertical tension. The objective function minimizes the weight of the PCRV. The constrained minimization problem is converted into an unconstrained problem by the addition of interior penalty functions to the objective function. The minimum is determined by the variable metric method (Davidson-Fletcher-Powell), using both values and derivatives of the modified objective function. The one-dimensional search is approximated by a method of Kund. Optimization variables are scaled. The method is applied to a pressure vessel like for THTR. It is found that the thickness of the cylindrical wall may be reduced considerably for the load cases considered in the optimization. The thickness of the cover is reduced slightly. The largest reduction in wall thickness occurs at the junction of wall and cover. (Auth.)

  14. Renovation of the sealing planes of WWER-400 reactors pressure vessel

    International Nuclear Information System (INIS)

    Jablonicky, P.; Pilat, P.

    2007-01-01

    An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

  15. Nuclear reactor fuel sub-assemblies

    International Nuclear Information System (INIS)

    Ford, J.; Bishop, J.F.W.

    1981-01-01

    An improved fuel sub-assembly for liquid metal cooled fast breeder nuclear reactors is described which facilitates dismantling operations for reprocessing purposes. The method of dismantling is described. (U.K.)

  16. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.E.

    1986-01-01

    A model has been developed to calculate the expansion and fragmentation of a corium jet, due to the evolution of dissolved gas, during a postulated core meltdown accident. Parametric calculations have been performed for a PWR high pressure accident scenario. Jet breakup occurs within a few jet diameters from the RPV. The diameter of the fragmented jet at the level of the reactor cavity floor is predicted to be 40-130 times the discharge diameter. Particles generated by fragmentation of corium melt are predicted to be in the 30-150 μm size range

  17. Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

    International Nuclear Information System (INIS)

    Ventura, Mirta A.; Rosso, Ricardo D.

    2002-01-01

    In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

  18. Standard technical specifications for Westinghouse pressurized water reactors (revision issued Fall 1981). Technical report

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1981-11-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

  19. The European Pressurized Water Reactor. A safe and competitive solution for future energy needs

    International Nuclear Information System (INIS)

    Leverenz, R.; Gerhard, L.; Goebel, A.

    2004-01-01

    The European Pressurized Water Reactor - the EPR - is a PWR in the 1600 MW class. Its design is based on experience feedback from several thousand reactors x years of light water reactor operation worldwide, primarily those incorporating the most recent technologies: the French N4 and the German KONVOI reactors. It is an evolutionary design that ensures continuity in the mastery of PWR technology, minimizing the risk for the customer. (author)

  20. An integrity evaluation method of the pressure vessel of nuclear reactors under pressurized thermal shock

    International Nuclear Information System (INIS)

    Matsubara, Masaaki; Okamura, Hiroyuki.

    1987-01-01

    Present paper proposes a new algorithm of the integrity evaluation of the pressure vessel of nuclear reactors under pressurized thermal shock, PTS. This method enables us to do an effective evaluation by superimposing proposed ''PTS state-transient curves'' and ''toughness transient curves'', and is superior to a conventional one in the following points; (1) easy to get an overall view of the result of PTS event for the variations of several parameters, (2) possible to evaluate a safety margin for irradiation embrittlement, and (3) enable to construct an Expert-friendly evaluation system. In addition, the paper shows that we can execute a safety assurance test by using a flat plate model with the same thickness as that of real plant. (author)

  1. Nuclear cleanup and decontamination for dismantling operations

    International Nuclear Information System (INIS)

    Bargues, S.; Solignac, Y.; Lapierre, Y.

    2003-01-01

    In the May 2003 issue of the review 'Controle', the French Nuclear Safety Authority (Autorite de Surete Nucleaire or ASN) reviewed the radiation protection and waste management principles applicable to dismantling operations carried out on nuclear installations, i.e. reactors, research laboratories, fuel cycle installations and nuclear power reactors. Estelle Chapelain, of the DGSNR (French General Directorate for Nuclear Safety and Radiation Protection), pointed out that dismantling work does not involve the same radioactive risks as operating an installation. For instance, 'the risk of disseminating radioactive material is generally greater because the dismantling process supposes the removal of one or more containment barriers'. In addition to this risk of internal exposure, the possibility of external irradiation of personnel must be taken into account due to the nature of the work carried out by the operators. The probability of conventional hazards is also accentuated, these hazards varying as work progresses (fire hazards during cutting operations, hazards associated with handling tasks, etc). Other risks must also be considered: hazards due to the ageing of installations, to loss of traceability, and finally the risks associated with waste management. Waste management falls within a strict regulatory framework specified by the decree dated December 31, 1999, which makes it compulsory to carry out a 'waste survey' with the aim of producing an inventory of waste and improving waste management. These surveys include 'waste zoning' to identify those areas liable to have been contaminated. These requirements lead operators to adapt their cleanup methodology in order to distinguish suspect rooms or equipment from those that can be deemed with certainty to be conventional. In its conclusion, the safety authority recalls the importance of 'the safety and radiation protection of dismantling operations being effectively managed and optimised, without imposing

  2. Embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1996-01-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

  3. Living PSA issues in France on pressurized water reactors

    International Nuclear Information System (INIS)

    Dewailly, J.; Deriot, S.; Dubreuil Chambardel, A.; Francois, P.; Magne, L.

    1993-09-01

    Two Probabilistic Safety Assessments (PSAs) carried out in France on 900 and 1300 MWe Pressurized Water Reactor units ended in 1990. These PSAs determined the core damage frequency for all plant operating conditions ranging from cold shutdown for refuelling to full power operation. Since 1990, these PSAs have been used increasingly as tools for applications such as accident precursor analysis, risk-based Technical Specifications, and maintenance optimization. In turn, these applications are used to enhance the initial PSAs. The notion of a ''living'' PSA which can be used and updated is slowly taking form. The accident precursor analysis consists in applying PSA event trees to obtain quick information on the potential consequences of a precursor event and on the corresponding probabilities of occurrence. A feedback on PSAs is provided by comparing them with actual operating incidents. The computation of the allowed outage time during power operation, based on the computerized models of Probabilistic Safety Assessments, requires adjustments: calculation of hourly risk of core damage under different reactor conditions without equipment unavailabilities. The proposed method also turns out to be an aid in determining the safe shutdown condition and procedure. Furthermore, when introducing a sufficient level of detail, PSA reliability models make it possible to compute contributions and to perform sensitivity studies in order to highlight those components for which a maintenance effort should be made. From the experience acquired up to now, there was felt to be a strong need to create guidelines for using PSAs that would simplify their implementation by the experts in charge of determining Technical Specifications, of maintenance programs, etc. who are not generally specialists in PSAs. For this purpose, it is necessary to improve the intelligibility of the models made in order for them to be used and to offer user's guides adapted to each application. Documents

  4. Reactor pressure vessel failure probability following through-wall cracks due to pressurized thermal shock events

    International Nuclear Information System (INIS)

    Simonen, F.A.; Garnich, M.R.; Simonen, E.P.; Bian, S.H.; Nomura, K.K.; Anderson, W.E.; Pedersen, L.T.

    1986-04-01

    A fracture mechanics model was developed at the Pacific Northwest Laboratory (PNL) to predict the behavior of a reactor pressure vessel following a through-wall crack that occurs during a pressurized thermal shock (PTS) event. This study, which contributed to a US Nuclear Regulatory Commission (NRC) program to study PTS risk, was coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory (ORNL). The PNL fracture mechanics model uses the critical transients and probabilities of through-wall cracks from the IPTS Program. The PNL model predicts the arrest, reinitiation, and direction of crack growth for a postulated through-wall crack and thereby predicts the mode of vessel failure. A Monte-Carlo type of computer code was written to predict the probabilities of the alternative failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. Plant-specific calculations were performed for the Oconee-1, Calvert Cliffs-1, and H.B. Robinson-2 reactor pressure vessels for the conditions of postulated transients. The model predicted that 50% or more of the through-wall axial cracks will turn to follow a circumferential weld. The predicted failure mode is a complete circumferential fracture of the vessel, which results in a potential vertically directed missile consisting of the upper head assembly. Missile arrest calculations for the three nuclear plants predict that such vertical missiles, as well as all potential horizontally directed fragmentation type missiles, will be confined to the vessel enclosre cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure modes that are most probable for postulated PTS events

  5. Construction management of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bohra, S.A.; Sharma, P.D.

    2006-01-01

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  6. The behavior of shallow flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Rolfe, S.T.

    1991-11-01

    Both analytical and experimental studies have shown that the effect of crack length, a, on the elastic-plastic toughness of structural steels is significant. The objective of this report is to recommend those research investigations that are necessary to understand the phenomenon of shallow behavior as it affects fracture toughness so that the results can be used properly in the structural margin assessment of reactor pressure vessels (RPVs) with flaws. Preliminary test results of A 533 B steel show an elevated crack-tip-opening displacement (CTOD) toughness similar to that observed for structural steels tested at the University of Kansas. Thus, the inherent resistance to fracture initiation of A 533 B steel with shallow flaws appears to be higher than that used in the current American Society of Mechanical Engineers (ASME) design curves based on testing fracture mechanics specimens with deep flaws. If this higher toughness of laboratory specimens with shallow flaws can be transferred to a higher resistance to failure in RPV design or analysis, then the actual margin of safety in nuclear vessels with shallow flaws would be greater than is currently assumed on the basis of deep-flaw test results. This elevation in toughness and greater resistance to fracture would be a very desirable situation, particularly for the pressurized-thermal shock (PTS) analysis in which shallow flaws are assumed to exist. Before any advantage can be taken of this possible increase in initiation toughness, numerous factors must be analyzed to ensure the transferability of the data. This report reviews those factors and makes recommendations of studies that are needed to assess the transferability of shallow-flaw toughness test results to the structural margin assessment of RPV with shallow flaws. 14 refs., 8 figs

  7. Underwater-manipulation system for measuring- and cutting tasks in dismantling decommissioned nuclear facilities. Final report

    International Nuclear Information System (INIS)

    Stegemann, D.; Reimche, W.; Hansch, M.; Spitzer, M.

    1995-01-01

    Not only manipulators are necessary for dismantling and inspection of structure parts in decomissioned nuclear facilities, but flexible underwater-vehicles. Free-diving underwater-vehicles for inspection and dismantling tasks are still not developed and tested. Aim of the project is the development of sensors and devices for the position determination and the depth regulation. For inspection tasks an ultrasonic measurement and dosimeter device shall be built up. A measurement device has been developed which evaluates the ultrasonic time of flight from a transmitter at the vehicle to several receivers, installed in the reactor pressure vessel. The depth regulation is based on a pressure sensor and the direct control of the thrusters. The ultrasonic measurements are realized by an adapted ultrasonic card, the γ-dosimetry with an ionization chamber and a pA-amplifier. An acoustic orientation system was built up, which measures very accurately with one transmitter mounted on the vehicle and four receivers. Problem occur by reflection from the walls of the basin. The depth regulation is working faultless. The ultrasonic device is preferably used for distance measurement. The radiation measurement device was tested and mounted in the vehicle. (orig./HP) [de

  8. A novel coordinated control for Integrated Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Zhao, Yuxin; Du, Xue; Xia, Genglei; Gao, Feng

    2015-01-01

    Highlights: • Proposed IPWR coordinated control strategy to avoid flow instability of OTSG. • Tuned PID controller parameters by Fuzzy kernel wavelet neural network with kernel trick and adaptive variable step-size. • Transition process exhibit the effectiveness of the novel IPWR control system. - Abstract: Integrated Pressurized Water Reactor (IPWR) has the characteristic of strong coupling, nonlinearity and complicated dynamic performance, which requires high standards of the control strategy and controller design. Most of IPWR systems utilize control strategy of ideal steady-state and PID controller, even though this strategy causes flow instability in the once through steam generator (OTSG) in low load conditions. Besides, the simple form of PID limits the performance developing which could not appropriately satisfy the requirements for quality. Motivated by these drawbacks, this paper proposes an IPWR coordinated control strategy and adopts PID controller to control each subsystem. The control strategy considers the system as a two-level hierarchical control system, and considers coordinating controller and bottom controllers. In the period of controller design, this strategy utilizes PID controller to control each subsystem, and modifies the controller parameters in real time by Fuzzy-KWNN algorithm, which adaptively achieves the system adjustment. Finally, simulation results are presented to exhibit the effectiveness of the proposed IPWR control system

  9. Advances in crack-arrest technology for reactor pressure vessels

    International Nuclear Information System (INIS)

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs

  10. Development of automatic ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik

    1994-08-01

    The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author)

  11. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs

  12. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  13. Updated embrittlement trend curve for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kirk, M.; Santos, C.; Eason, E.; Wright, J.; Odette, G.R.

    2003-01-01

    The reactor pressure vessels of commercial nuclear power plants are subject to embrittlement due to exposure to high energy neutrons from the core. Irradiation embrittlement of RPV belt-line materials is currently evaluated using US Regulatory Guide 1.99 Revision 2 (RG 1.99 Rev 2), which presents methods for estimating the Charpy transition temperature shift (ΔT30) at 30 ft-lb (41 J) and the drop in Charpy upper shelf energy (ΔUSE). A more recent embrittlement model, based on a broader database and more recent research results, is presented in NUREG/CR-6551. The objective of this paper is to describe the most recent update to the embrittlement model in NUREG/CR-6551, based upon additional data and increased understanding of embrittlement mechanisms. The updated ΔT30 and USE models include fluence, copper, nickel, phosphorous content, and product form; the ΔT30 model also includes coolant temperature, irradiation time (or flux), and a long-time term. The models were developed using multi-variable surface fitting techniques, understanding of the ΔT30 mechanisms, and engineering judgment. The updated ΔT30 model reduces scatter significantly relative to RG 1.99 Rev 2 on the currently available database for plates, forgings, and welds. This updated embrittlement trend curve will form the basis of revision 3 to Regulatory Guide 1.99. (author)

  14. Development of automatic ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author).

  15. Remerschen nuclear power station with BBR pressurized water reactor

    International Nuclear Information System (INIS)

    Hoffmann, J.P.

    1975-01-01

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

  16. Embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1998-01-01

    The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes. (orig.)

  17. Shallow-crack toughness results for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Theiss, T.J.; Shum, D.K.M.; Rolfe, S.T.

    1992-01-01

    The Heavy Section Steel Technology Program (HSST) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. To complete this investigation, techniques were developed to determine the fracture toughness from shallow-crack specimens. A total of 38 deep and shallow-crack tests have been performed on beam specimens about 100 mm deep loaded in 3-point bending. Two crack depths (a ∼ 50 and 9 mm) and three beam thicknesses (B ∼ 50, 100, and 150 mm) have been tested. Techniques were developed to estimate the toughness in terms of both the J-integral and crack-tip opening displacement (CTOD). Analytical J-integral results were consistent with experimental J-integral results, confirming the validity of the J-estimation schemes used and the effect of flaw depth on fracture toughness. Test results indicate a significant increase in the fracture toughness associated with the shallow flaw specimens in the lower transition region compared to the deep-crack fracture toughness. There is, however, little or no difference in toughness on the lower shelf where linear-elastic conditions exist for specimens with either deep or shallow flaws. The increase in shallow-flaw toughness compared with deep-flaw results appears to be well characterized by a temperature shift of 35 degree C

  18. Method of reactivity control in pressure tube reactor

    International Nuclear Information System (INIS)

    Fukumura, Nobuo.

    1988-01-01

    Purpose: To provide a method of controlling reactivity in a pressure tube reactor at high conversion ratio intended for high burn-up degree. Method: Control tubes are inserted in heavy water moderator. Light water is filled in the tubes at the initial burning stage. Along with the advance of the burning, the light water is gradually removed and replaced with gases of less reactive nuclear reactivity with neutrons such as air or gaseous carbon dioxide. The tubes are made of less neutron absorbing material such as aluminum. By filling light water, infinite multiplication factor is reduced to suppress the reactivity at the initial burning stage. As light water is gradually removed and replaced with air, etc., it provides an effect like that elimination of heavy water moderator to increase the conversion ratio. Accordingly, nuclear fission materials are produced additionally by so much to extend the burn-up degree. In this way, it can provide excellent effect in realizing high burn-up ratio and high conversion ratio. (Kamimura, M.)

  19. Numerical study of optimal equilibrium cycles for pressurized water reactors

    International Nuclear Information System (INIS)

    Mahlers, Y.P.

    2003-01-01

    An algorithm based on simulated annealing and successive linear programming is applied to solve equilibrium cycle optimization problems for pressurized water reactors. In these problems, the core reload scheme is represented by discrete variables, while the cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are treated as continuous variables. The enrichments are considered to be distinct in all feed fuel assemblies. The number of batches and their sizes are not fixed and also determined by the algorithm. An important feature of the algorithm is that all the parameters are determined by the solution of one optimization problem including both discrete and continuous variables. To search for the best reload scheme, simulated annealing is used. The optimum cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are determined for each reload pattern examined using successive linear programming. Numerical results of equilibrium cycle optimization for various values of the effective price of electricity and fuel reprocessing cost are studied

  20. Software to study the control strategy of pressurized water reactor

    International Nuclear Information System (INIS)

    Oliveira, Jose Ricardo de

    2002-01-01

    The computational program, result of this work, is a tool developed for the study of the control of Pressurized Water Reactors (PWR) constituted by only one coolant loop. The implementation of a user friendly interface for input/output data, makes the program also suitable for training and teaching applications. As design premise, it was considered enough just the modeling of the primary circuit, using as interface with the secondary circuit, a simplified differential equation of the temperature associated with the secondary power. All the incorporated dynamic equations to the model were developed using basic laws of conservation, boundary conditions and hypotheses appropriated to the control study. To arrive to the final model, core thermal and hydraulic characteristics and design data were obtained from of the available bibliography and adapted for a conceptual peculiar design of a small PWR. The whole program and all input/output interfaces were developed using the software Matlab, version 5.L Sub-routines of numeric integration based on the Runge-Kutta 4 method were applied, to solve the set of ordinary differential equations. (author)

  1. Different approaches to estimation of reactor pressure vessel material embrittlement

    Directory of Open Access Journals (Sweden)

    V. M. Revka

    2013-03-01

    Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

  2. Design description of the European pressurized water reactor

    International Nuclear Information System (INIS)

    Leverenz, R.

    1999-01-01

    The EPR (the European Pressurized Water Reactor) is an evolutionary PWR developed by Nuclear Power International and its parent companies, Framatome and Siemens, in co-operation with Electricite de France and German Utilities. NPI can rely on the huge experience gained by its parent companies; they have constructed more than 100 nuclear power plants throughout the world. The total installed capacity exceeds 100,000 MW - about 25% of the total world-wide figure. Following the conceptual design phase of the so-called Common Product conducted by NPI, Framatome and Siemens, from 1989 through 1991, Electricite de France (EDF) and several major German utilities decided to merge their own development programmes, - the N4 Plus and REP 2000 projects on the French side and the further development of the KONVOI technology on the German side, - with the NPI project. From that time on, the NPI project became one single common development line for both countries. In parallel, EDF and the German utilities decided to establish, together with other European utilities, specifications that would represent common utility views on the design and performance of future nuclear power plants. These are documented in the European Utility Requirements (EURs). The basic design has been completed in 1997, and in 1998 a design optimization is being carried out with the goal to even increase the economic competitiveness of nuclear power. This paper provides a brief design description of the EPR. (author)

  3. Revisiting the reactor pressure vessel for long-time operation

    International Nuclear Information System (INIS)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-01-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  4. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Bertram, W.

    1975-01-01

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  5. Fibre optic sensors in pressurized water reactor alternators

    International Nuclear Information System (INIS)

    Favennec, J.M.; Piguet, M.

    1994-01-01

    Measurement in the electrical engine environment (alternator, transformer...) is identified as one of the two main applications of fibre optic sensors within EDF; the other application niche is the monitoring of civil works (dams, containment building of nuclear reactors...). At the EDF Research and Development Division, temperature and vibration fibre optic sensors were evaluated by the Metrology Service, since their use is under consideration for alternator monitoring. For alternator stator thermal monitoring, the BERTIN company developed a fibre optic sensor network. The optic coding technique is based on broadband source spectral modulation; the sensors are interrogated sequentially by electronic commutation. For alternator stator vibration monitoring, a fibre optic accelerometer was developed in the frame of a manufacturers and universities consortium supported by the French Research and Technology Ministry. The accelerometer is of cantilever beam type and its networking is possible by chromatic multiplexing. The Metrology Service evaluated these temperature and vibration sensors in order to verify their metrological characteristics (bias error, hysteresis, repeatability, resolution, noise, amplitude linearity, response time, frequency response, etc.) and to test their behaviour in harsh alternator environmental conditions (pressure, vibrations and temperature). Ageing and accidental condition resistance tests were also carried out. Temperature sensor test results were very satisfactory. An eight-sensor BERTIN prototype was installed on the Tricastin 1 alternator during the september 1993 nuclear station periodic stop. On the contrary, the accelerometers presented deficient metrological characteristics (shorter span than foreseen, low repeatability...). They need some improvements and could not be installed on alternators. (authors). 5 refs., 8 figs

  6. Apparatus for localizing disturbances in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Sykora, D.

    1989-01-01

    The invention according to CS-PS 177386, entitled ''Apparatus for increasing the efficiency and passivity of the functioning of a bubbling-vacuum system for localizing disturbances in nuclear power plants with a pressurized water reactor'', concerns an important area of nuclear power engineering that is being developed in the RGW member countries. The invention solves the problems of increasing the reliability and intensification during the operation of the above very important system for guaranteeing the safety of the standard nuclear power plants of Soviet design. The essence of the invention consists in the installation of a simple passively operating supplementary apparatus. Consequently, the following can be observed in the system: first an improvement and simultaneous increase in the reliability of its function during the critical transition period, which follows the filling of the second space with air from the first space; secondly, elimination of the hitherto unavoidable initiating role of the active sprinkler-condensation device present; thirdly, a more effective performance and subjection of the elements to disintegration of the water flowing from the bubbling condenser into the first space; and fourthly, an enhanced utilization of the heat-conducting ability of the water reservoir of the bubbling condenser. Representatives of the supplementary apparatus are autonomous and local secondary systems of the sprinkler-sprayer without an insert, which spray the water under the effect of gravity. 1 fig

  7. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  8. Recent development for inservice inspection of reactor pressure vessels

    International Nuclear Information System (INIS)

    Fischer, K.; Engl, G.; Rathgeb, W.; Heumueller, R.

    1991-01-01

    The German Nuclear Code (KTA-rules) requires a full scope inservice inspection (ISI) of reactor pressure vessels within a period of four years. This has a remarkable influence on plant operation and economy. Therefore, the development of advanced inspection equipment and techniques is directed not only to the enhancement of defect detectability and flaw sizing capabilities but also to reducing inspection times. A new manipulator system for PWR vessels together with fast data processing reduces the time for ISI of modern RPVs to 7 days. A new multichannel UT-system based on ALOK principle offers increased ultrasonic information with comfortable and rapid evaluation and presentation of results together with enhanced sizing capabilities. For specific inspection problems characterized by geometrical complexity the application of phased array probes in connection with UT-tomography provides improved ultrasonic information together with a streamlined manipulator principle and simplification of set up and tear down at the component which results in considerable reduction of radiation exposure. (orig.)

  9. Contributions to the energetical role of advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1984-06-01

    Three articles written by the author in the past two years and all concerned with energy production of the Advanced Pressurized Water Reactor (APWR) are collated to form this report. The standpoint is made that the APWR using mixed oxide-fuel (about 8% Pusub(fiss) and Usub(depl)) would entail similar capital costs as a PWR, but with conversion rates of 0.85< C<0.95. Given this assumption the analysis shows that the high specific plutonium inventories (being proportional to the conversion ratio in the interested range) result in a strong damping effect on the growth rate, the effective uranium utilization, and on the total nuclear power level. Over one century an APWR strategy is 'far from equilibrium' and to describe this dynamic situation using static APWR data is not appropriate. If nuclear fission energy is to play a substantial and not just a marginal role in providing energy for the future (in the region of 5 - 8 TWsub(el) in 50 years time, corresponding to a share of 20 - 25% of the total world demand), clearly the fast breeder strategy offers the better chance of achieving the goal. (Auth.)

  10. Measurement and analysis of pressure tube elongation in the Douglas Point reactor

    International Nuclear Information System (INIS)

    Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

    1980-02-01

    Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

  11. CFD simulation analysis and validation for CPR1000 pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Mingqian; Ran Xiaobing; Liu Yanwu; Yu Xiaolei; Zhu Mingli

    2013-01-01

    Background: With the rapid growth in the non-nuclear area for industrial use of Computational fluid dynamics (CFD) which has been accompanied by dramatically enhanced computing power, the application of CFD methods to problems relating to Nuclear Reactor Safety (NRS) is rapidly accelerating. Existing research data have shown that CFD methods could predict accurately the pressure field and the flow repartition in reactor lower plenum. But simulations for the full domain of the reactor have not been reported so far. Purpose: The aim is to determine the capabilities of the codes to model accurately the physical phenomena which occur in the full reactor vessel. Methods: The flow field of the CPR1000 reactor which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX. The pressure loss in reactor pressure vessel, the hydraulic loads of guide tubes and support columns, and the bypass flow of head dome were obtained by calculations for the full domain of the reactor. The results were validated by comparing with the determined reference value of the operating nuclear plant (LingAo nuclear plant), and the transient simulation was conducted in order to better understand the flow in reactor pressure vessel. Results: It was shown that the predicted pressure loss with CFD code was slightly different with the determined value (10% relative deviation for the total pressure loss), the hydraulic loads were less than the determined value with maximum relative deviation 50%, and bypass flow of head dome was approximately the same with determined value. Conclusion: This analysis practice predicts accurately the physical phenomena which occur in the full reactor vessel, and can be taken as a guidance for the nuclear plant design development and improve our understanding of reactor flow phenomena. (authors)

  12. Improvement of methods to evaluate brittle failure resistance of the WWER reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Popov, A A; Parshutin, E V [Engineering Center of Nuclear Equipment Strength, Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Rogov, M F; Dragunov, U G [Experimenter` s and Designer` s Office ` ` Hydropress` ` (Russian Federation)

    1997-09-01

    At the next 10 years a number of Russian WWER nuclear power plants will complete its design lifetime. Normative methods to evaluate brittle failure resistance of the reactor pressure vessels used in Russia have been intended for design stage. The evaluation of reactor pressure vessel lifetime in operation stage demands to create new methods of calculation and new methods for experimental evaluation of brittle failure resistance degradation. The main objective of the study in this type of reactor is weldment number 4. In this report an analysis is made of methods to determine critical temperature of reactor materials including the results of instrumented Charpy testing. 12 figs.

  13. Status of reactor pressure vessel embrittlement study in Japan

    International Nuclear Information System (INIS)

    Sasajima, H.

    1997-01-01

    Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

  14. Guide to the periodic inspection of nuclear reactor steel pressure vessels

    International Nuclear Information System (INIS)

    1969-01-01

    This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

  15. Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel

    International Nuclear Information System (INIS)

    Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A.

    1993-08-01

    The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117

  16. FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

    International Nuclear Information System (INIS)

    Dickson, T.L.

    1993-01-01

    This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness

  17. Activities in the Czech Republic for reactor pressure components lifetime management

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1994-01-01

    The following activities in the Czech republic for reactor pressure components lifetime management are described: upgrading and safety assurance of nuclear power plants (NPP) with reactors of WWER-440/V-230 type, safety assurance of NPPs with reactors of WWER-440/V-213, lifetime management programme of NPPs with WWER-440/V-213 reactors, preparation of start-up of NPPs with WWER-1000/V-320 reactors, preparation of guides for lifetime as well as defect allowability evaluation in main components of primary and secondary circuits. 3 figs

  18. Method of cooling a pressure tube type reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro.

    1983-01-01

    Purpose: To improve the operation efficiency of a nuclear reactor by carrying out cooling depending on the power distribution in the reactor core. Constitution: Reactor core channels are divided into a plurality of channel groups depending on the reactor power, and a water drum and a pump are disposed to each of the channel groups so as to increase the amount of coolants in response to the magnitude of the power from each of the channel groups. In this way, the minimum limiting power ratio can be increased. (Seki, T.)

  19. Importance of low-level radioactive wastes in dismantling strategy in CEA (FRANCE)

    International Nuclear Information System (INIS)

    Lafaille, C.

    1991-01-01

    This paper describes the advance used in C.E.A. to realize dismantling operations in the best technical and economical conditions. Particularly, for low-level radioactive waste management CEA's advance defines, first, the final destination of dismantling materials: - recycling in public lands for level activity inferior to 1 Bq/g; directly or after transformation (melting, calcination, extrusion) - storage in a ground disposal, after compacting, encapsulation or drumming. Two examples are given: - Marcoule G2 - G3 reactor dismantling - Gaseous diffusion plants demolition (COGEMA Pierrelatte)

  20. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)