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Sample records for sodium cooled fbr

  1. A new concept of hydrogen production system for sodium cooled FBR

    International Nuclear Information System (INIS)

    Nakagiri, Toshio; Aoto, Kazumi; Hoshiya, Taiji

    2004-01-01

    A new thermo-chemical and electrolytic hybrid hydrogen production process (thermo-chemical and electrolytic Hybrid Hydrogen process in Lower Temperature range: HHLT) is newly proposed by the Japan Nuclear Cycle Development Institute (JNC) to realize the hydrogen production from water by using the heat generation of sodium cooled Fast Breeding Reactor (FBR). The HHLT process is based on the sulfuric acid (H 2 SO 4 ) synthesis and decomposition processes developed earlier (Westinghouse process), and sulfur trioxide (SO 3 ) decomposition process of HHLT is facilitated by electrolysis with ionic oxygen conductive solid electrolyte to reduce operating temperature 200degC-300degC lower than Westinghouse process. Decomposition processes of SO 3 were confirmed with the cell voltage lower than 0.5 V at 500degC-600degC using 8mol yttria stabilized zirconia (8molYSZ) solid electrolyte and platinum electrode. Therefore, total voltage required for HHLT is expected to be lower than 1.0 V, because the voltage required for sulfuric acid synthesis is about 0.5V. Thermal efficiency of HHLT based on chemical reactions was roughly estimated to be within the range of 35% to 55% under the influence of H 2 SO 4 concentration and heat recovery. These results show the possibility of development of a new hydrogen production process which needs low splitting voltage and has high efficiency at around 500degC, utilizing the heat generation of sodium cooled FBR. SO 3 splitting with the voltage lower than 0.5V was confirmed at about 500degC experimentally, and ideal thermal efficiency of the cycle based on chemical reactions was evaluated. Furthermore, test apparatus to substantiate whole process of HHLT was manufactured. (author)

  2. Practicalization strategic research of FBR cycle

    International Nuclear Information System (INIS)

    2000-01-01

    Practicalization strategic research of FBR cycle consists of two phases such as phase I (FY 1999-2000) and phase II (to FY 2005). In every phase, research and development plants and results are checked and reviewed. The assessment indexes are five development objects such as safety, economical efficiency, resource effective utilization, environmental load decrease and nuclear non-proliferation and technical realization, too. Reactor core, FBR plant system and fuel cycle system are investigated. We selected the research subjects of cooling materials as sodium, heavy metals (lead and lead bismuth alloy), gas (carbon dioxide and helium) and water (boiling water, power water and supercritical pressure water) and fuel types as cladding tube fuel (oxide, nitride and metal) and coated fuel particle (oxide and nitride) for helium gas cooling reactor. In FY1999, the good reactor core and FBR plant system for every cooling materials are studied. Two reprocessing (a wet reprocessing using aqueous solution and a dry method) were selected. In FY 2000, we will investigate effects of throughput, plant concept and cost and evaluate achievement of development objects and then decide the development plan. (S.Y.)

  3. Conceptual design study on simplified and safer cooling systems for sodium cooled FBRs

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Shimakawa, Yoshio; Ishikawa, Hiroyasu; Kubota, Kenichi; Kobayashi, Jun; Kasai, Shigeo

    2000-06-01

    The objective of this study is to create the FBR plant concepts increasing economy and safety for the Phase-I 'Feasibility Studies on Commercialized Fast Reactor System'. In this study, various concepts of simplified 2ry cooling system for sodium cooled FBRs are considered and evaluated from the view points of technological feasibility, economy, and safety. The concepts in the study are considered on the basis of the following points of view. 1. To simplify 2ry cooling system by moderating and localizing the sodium-water reaction in the steam generator of the FBRs. 2. To simplify 2ry cooling system by eliminating the sodium-water reaction using integrated IHX-SG unit. 3. To simplify 2ry cooling system by eliminating the sodium-water reaction using a power generating system other than the steam generator. As the result of the study, 12 concepts and 3 innovative concepts are proposed. The evaluation study for those concepts shows the following technical prospects. 1. 2 concepts of integrated IHX-SG unit can eliminate the sodium-water reaction. Separated IHX and SG tubes unit using Lead-Bismuth as the heat transfer medium. Integrated IHX-SG unit using copper as the heat transfer medium. 2. Cost reduction effect by simplified 2ry cooling system using integrated IHX-SG unit is estimated 0 to 5%. 3. All of the integrated IHX-SG unit concepts have more weight and larger size than conventional steam generator unit. The weight of the unit during transporting and lifting would limit capacity of heat transfer system. These evaluation results will be compared with the results in JFY 2000 and used for the Phase-II study. (author)

  4. Heavy liquid metal cooled FBR. Results 2001

    International Nuclear Information System (INIS)

    Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu

    2003-08-01

    In the feasibility studies of commercialization of an FBR fuel cycle system, the targets are economical competitiveness to future LWRs, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation, besides ensuring safety. Both medium size pool-type lead-bismuth cooled reactor with primary pumps system and without primary pumps system are studied to pursue their improvement in heavy metal coolant considering design requirements form plant structures. The design of plant systems are reformed, and the conceptual design is made and the commodities are analyzed. (1) Conceptual design of lead-bismuth cooled reactor with pumping system: Electrical output 750 MWe and 4-module system. The heat-mass balance is optimized and drawings are made about plant layout, cooling system, reactor structure and cooling component structures. (2) Structural analysis of main components. (3) Conceptual design of natural circulation type lead-bismuth cooled reactor: Electrical output 550 MWe and 6-module system. The heat-mass balance is optimized and drawings are made about plant layout, cooling system, reactor structure and cooling component structures. (4) Study of R and D program. (author)

  5. FBR structural material test facility in flowing sodium environment

    International Nuclear Information System (INIS)

    Shanmugasundaram, M.; Kumar, Hemant; Ravi, S.

    2016-01-01

    In Fast Breeder Reactor (FBR), components such as Control and Safety Rod Drive Mechanism (CSRDM), Diverse Safety Rod Drive Mechanism (DSRDM), Transfer arm and primary sodium pumps etc., are experiencing friction and wear between the moving parts in contact with liquid sodium at high temperature. Hence, it is essential to evaluate the friction and wear behaviour to validate the design of components. In addition, the above core structural reactor components such as core cover plate, control plugs etc., undergoes thermal striping which is random thermal cycling induced by flow stream resulting from the mixing of non isothermal jets near that component. This leads to development of surface cracks and assist in crack growth which in turn may lead to failure of the structural component. Further, high temperature components are often subjected to low cycle fatigue due to temperature gradient induced cyclic thermal stresses caused by start-ups, shutdowns and transients. Also steady state operation at elevated temperature introduces creep and the combination of creep and fatigue leads to creep-fatigue interactions. Therefore, resistance to low cycle fatigue, creep and creep-fatigue are important considerations in the design of FBR components. Liquid sodium is used as coolant and hence the study of the above properties in dynamic sodium are equally important. In view of the above, facility for materials testing in sodium (INSOT) has been constructed and in operation for conducting the experiments such as tribology, thermal stripping, low cycle fatigue, creep and creep-fatigue interaction etc. The salient features of the operation and maintenance of creep and fatigue loops of INSOT facility are discussed in detail. (author)

  6. Mass transfer of steels for FBR in sodium loop

    International Nuclear Information System (INIS)

    Susukida, Hiroshi; Yonezawa, Toshio; Ueda, Mitsuo; Imazu, Takayuki; Kiyokawa, Teruyuki.

    1976-06-01

    In order to grasp quantitatively the corrosion and mass transfer of steels for FBR in sodium loop and to establish their allowable stress value and corrosion rate, a special sodium loop for material testing was designed and fabricated and the steels were given 3010 hours exposing test in the sodium loop. This paper gives the outline of the sodium loop and the results of the test. (1) Carburization and a slight increase in weight were observed in the specimens of type 304 stainless steel exposed in the sodium loop for 3010 hours, while decarburization was observed in the specimens of 2 1/4 Cr-1 Mo steel. It is considered that these phenomena were caused by the downstream factor of the sodium loop. (2) A remarkable decrease of Charpy absorbed energy was observed in the specimens of type 304 stainless steel exposed in the sodium loop. It is considered that this resulted from the weakening of the grain boundary due to heat history and mass transfer. (3) The specimens exposed in the sodium loop must be washed by ultrasonic waves in a water bath after washing in alcohol. (auth.)

  7. Study on thermal electric conversion system for FBR plant. Investigation for effective EVST waste heat recovery system

    International Nuclear Information System (INIS)

    Maekawa, Isamu; Kurata, Chikatoshi

    2004-02-01

    Recently, it has been important to reuse discharged heat energy from present nuclear plant, especially from sodium cooled FBR, which are typical high temperature system, in the view of reduction of environmental burden and improvement of heat efficiency for plant. The thermal electric conversion system can work only the temperature difference and has been applied to the limited fields such as space or military, however, that results show good merits for reliability, maintenance free, and so on. Recently, the development of new thermal electric conversion elements has made remarkable progress. In this study, for the effective utilization of waste heat from Monju', the prototype plant of FBR, we made an investigation of electric power generating system maintaining the cooling faculty by applying the thermal electric conversion system to sodium cooling line of EVST. Using the new type iron based thermal electric conversion elements, which are plentiful, economical and good for environmental harmonization, we have calculated the amount of heat exchange and power generation from sodium cooling line of EVST, and have investigated the module sizing, cost and subject to be settled. The results were , (1)The amount of power generation from sodium cooling line of EVST is smaller about one figure than motive power of sodium cooler fan. However, if Seebeck coefficient and heat conductivity of iron based thermal electric conversion elements shall be improved, power from sodium cooling line shall be able to cover the motive power. (2) The amount of heat released from sodium cooling line after the installation of thermal electric conversion module covers the necessity to maintain the sodium cooling faculty. (3) In case of the installation of module to the sodium cooler, it should be reconstructed because of tube arrangement modification. In case of the installation of module to the sodium connecting line, air ventilation system is needed to suppress the room temperature. (4) As

  8. Development of hydrogen production technology using FBR

    International Nuclear Information System (INIS)

    Ono, Kiyoshi; Otaki, Akira; Chikazawa, Yoshitaka; Nakagiri, Toshio; Sato, Hiroyuki; Sekine, Takashi; Ooka, Makoto

    2004-06-01

    This report describes the features of technology, the schedule and the organization for the research and development regarding the hydrogen production technology using FBR thermal energy. Now, the hydrogen production system is proposed as one of new business models for FBR deployment. This system is the production of hydrogen either thermal energy at approximately from 500degC to 550degC or electricity produced by a sodium cooled FBR. Hydrogen is expected to be one of the future clean secondary energies without carbon-dioxide emission. Meanwhile the global energy demand will increase, especially in Asian countries, and the energy supply by fossil fuels is not the best choice considering the green house effect and the stability of energy supply. The development of the hydrogen technology using FBR that satisfies 'sustainable energy development' and 'utilization of energies free from environmental pollution' will be one of the promising options. Based on the above mentioned recognition, we propose the direction of the development, the issues to be solved, the time schedule, the budget, and the organization for R and D of three hydrogen production technologies, the thermochemical hybrid process, the low temperature steam reforming process, and the high temperature steam electrolysis process in JNC. (author)

  9. Toward commercialization of FBR cycle (1). Promotion of R and D on technologies maintaining sustainable society

    International Nuclear Information System (INIS)

    Nagaoki, Yoshihiro; Nagura, Fuminori; Sakaguchi, Tomoyoshi; Kawasaki, Hirotsugu; Kikuchi, Shin

    2008-01-01

    The FBR cycle is a key technology maintaining a sustainable society through efficient utilization of limited uranium resources and conformance to global environmental protection. The domestic and overseas R and D of the FBR cycle entered on a new phase aiming at its commercialization and JAEA started the Fast Reactor Cycle Technology Development (FaCT) project. The FaCT project targeted at international standardization of the FBR fuel cycle and promoting the advanced R and D on the innovative technologies to increase cost-efficiency and reliability for the commercialization under international competition and cooperation. The combination of a sodium cooled FBR and advanced fuel cycle system with advanced aqueous reprocessing and simplified pelletizing fuel fabrication was selected a major concept. (T. Tanaka)

  10. Design study on sodium cooled large-scale reactor

    International Nuclear Information System (INIS)

    Murakami, Tsutomu; Hishida, Masahiko; Kisohara, Naoyuki

    2004-07-01

    In Phase 1 of the 'Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2, design improvement for further cost reduction of establishment of the plant concept has been performed. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2003, which is the third year of Phase 2. In the JFY2003 design study, critical subjects related to safety, structural integrity and thermal hydraulics which found in the last fiscal year has been examined and the plant concept has been modified. Furthermore, fundamental specifications of main systems and components have been set and economy has been evaluated. In addition, as the interim evaluation of the candidate concept of the FBR fuel cycle is to be conducted, cost effectiveness and achievability for the development goal were evaluated and the data of the three large-scale reactor candidate concepts were prepared. As a results of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  11. The Swiss contribution to the FBR development

    Energy Technology Data Exchange (ETDEWEB)

    Hudina, M [Institut Federal de Recherches en Matiere de Reacteurs, Wuerenlingen (Switzerland); and others

    1981-05-01

    The program of fast breeder reactors development in Switzerland is considered from two points of view: energy self-sufficiency and optimization of fuel cycle. Research and development program covers: safety features of LMFBRs, development of mixed carbide fuel elements, study of steam generators transient behaviour, influence of various cooling concepts on thermal efficiency, techniques for detecting cover gas bubbles in the primary sodium circuit. This paper includes cost od the research and development activities as well as description of the future aims of the FBR projects.

  12. The Swiss contribution to the FBR development

    International Nuclear Information System (INIS)

    Hudina, M.

    1981-01-01

    The program of fast breeder reactors development in Switzerland is considered from two points of view: energy self-sufficiency and optimization of fuel cycle. Research and development program covers: safety features of LMFBRs, development of mixed carbide fuel elements, study of steam generators transient behaviour, influence of various cooling concepts on thermal efficiency, techniques for detecting cover gas bubbles in the primary sodium circuit. This paper includes cost od the research and development activities as well as description of the future aims of the FBR projects

  13. Study on hydrogen production using the fast breeder reactors (FBR)

    International Nuclear Information System (INIS)

    Kani, Yoshio

    2003-01-01

    As the fast breeder reactor (FBR) can effectively convert uranium-238 difficult to carry out nuclear fission at thermal neutron reactors to nuclear fissionable plutonium-239 to use it remarkable upgrading of application on uranium can be performed, to be expected for sustainable energy source. And, by reuse minor actinides of long half-life nuclides in reprocessed high level wasted solutions for fuels of nuclear reactors, reduction of radioactive poison based on high level radioactive wastes was enabled. As high temperature of about 800 centigrade was required on conventional hydrogen production, by new hydrogen production technique even at operation temperature of sodium-cooled FBR it can be enabled. Here were described for new hydrogen production methods applicable to FBR on palladium membrane hydrogen separation method carrying out natural gas/steam modification at reaction temperature of about 500 centigrade, low temperature thermo-chemical method expectable simultaneous simplification of production process, and electrolysis method expected on power load balancing. (G.K.)

  14. Development of FBR technology in the FBR 'Joyo'

    International Nuclear Information System (INIS)

    Nara, Yoshihiko; Akiyama, Takao; Sato, Isao; Mizoo, Nobutatsu; Yoshimi, Hirotaka; Shimada, Takashi

    1986-01-01

    Power Reactor and Nuclear Fuel Development Corp. has advanced the construction of the prototype FBR ''Monju'', and the ground breaking ceremony was held on October 28, 1985. For the design and construction of Monju, the experience, achievement, and the results of development by the own effort and international cooperation gained by the experimental FBR ''Joyo'' have been reflected. It is important to develop the core management technology, operation-supporting system, the techniques of regular inspection, maintenance and repair, the reduction of radiation exposure and so on, to accumulate the experience, and to reflect those accurately to Monju. The operation history of the experimental FBR ''Joyo'', the international joint research on FBRs using the Joyo, the results regarding the characteristic technology of FBRs such as the reactor core, fuel and control rods, sodium technology, the construction of machinery and equipment, and the plant system the plan of developing the high grade technology of FBRs such as the development of fuel and materials, the improvement of reliability and the development of operation management techniques, the verifying test of new technology such as spent fuel storage, the new system for sodium purification and the techniques for analyzing earthquake response, and the international cooperation are reported. (Kako, I.)

  15. 3-D numerical simulation on the vibration of liquid sodium's free surface in sodium pool of FBR

    International Nuclear Information System (INIS)

    Han Biao; Yao Zhaohui; Ye Hongkai; Wang Xuefang

    1997-01-01

    This paper succeeds in simulating three-dimensional incompressible flows with free surface, complicated in-flow and out-flow boundary conditions and internal obstacles, and also can treat these fluid flows in arbitrary shape vessel using a partial cell. According to all kinds of the element influencing the free surface's vibration in sodium pool it may give the various wave's form, the highest and lowest position, and the amount of the vibration. This paper introduces the brief principle of VOF numerical method, develops the computational program based on NASA-VOF3D, provides some results about the free surface's vibration in sodium pool of FBR

  16. Leak detector for a steam generator in FBR type reactors

    International Nuclear Information System (INIS)

    Miyaji, Nobuyoshi.

    1979-01-01

    Purpose: To facilitate maintenance for liquid leak detectors such as exchange of nickel membrane sensors during operation in a sodium-cooled fbr type reactor. Constitution: A pipeway capable of supplying a cover gas such as argon into the cylinder of a hydrogen detector containing a nickel membrane sensor is provided in a liquid leak detector constituting a part of a by-pass loop. The pipeway is also adapted to be evacuated. A pipeway and a small sodium tank for drain use are provided on the side of the by-pass loop near valves. Then, after closing the inlet and outlet valves to disconnect the by-pass loop from the sodium main pipeway, the cover gas is supplied to drive liquid sodium to the drain tank. After the drain of the liquid sodium, the sensor can be replaced. (Ikeda, J.)

  17. Sodium Exposure Tests on Limestone Concrete Used as Sacrificial Protection Layer in FBR

    International Nuclear Information System (INIS)

    Parida, F.C.; Das, S.K.; Sharma, A.K.; Rao, P.M.; Ramesh, S.S.; Somayajulu, P.A.; Malarvizhi, B.; Kasinathan, N.

    2006-01-01

    Hot sodium coming in contact with structural concrete in case of sodium leak in FBR system cause damage as a result of thermo-chemical attack by burning sodium. In addition, release of free and bound water from concrete leads to generation of hydrogen gas, which is explosive in nature. Hence limestone concrete, as sacrificial layer on the structural concrete in FBR, needs to be qualified. Four concrete blocks of dimension 600 mm x 600 mm x 300 mm with 300 mm x 300 mm x 150 mm cavity were cast and subjected to controlled sodium exposure tests. They have composition of ordinary portland cement, water, fine and coarse aggregate of limestone in the ratio of 1: 0.58: 2.547: 3.817. These blocks were subjected to preliminary inspection by ultrasonic pulse velocity technique and rebound hammer tests. Each block was exposed for 30 minutes to about 12 kg of liquid sodium (∼ 120 mm liquid column) at 550 deg. C in open air, after which sodium was sucked back from the cavity of the concrete block into a sodium tank. On-line temperature monitoring was carried out at strategic locations of sodium pool and concrete block. After removing sodium from the cavity and cleaning the surfaces, rebound hammer testing was carried out on each concrete block at the same locations where data were taken earlier at pre-exposed stage. The statistical analysis of rebound hammer data revealed that one of the concrete block alone has undergone damage to the extent of 16%. The loss of mass occurred for all the four blocks varied from 0.6 to 2.4% due to release of water during the test duration. Chemical analysis of sodium in concrete samples collected from cavity floor of each block helped in generation of depth profiles of sodium monoxide concentration for each block. From this it is concluded that a bulk penetration of sodium up to 30 mm depth has taken place. However it was also observed that at few local spots, sodium penetrated into concrete up to 50 mm. Cylindrical core samples of 50 mm x 150

  18. The report of inspection and repair technology of sodium cooled reactors

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Uchita, Masato; Konomura, Mamoru

    2002-12-01

    Sodium is the most promising candidate of an FBR coolant because of its excellent properties such as high thermal conductivity. Whereas, sodium reacts with water/air and its opaqueness makes it difficult to inspect sodium components. These weaknesses of sodium affect not only plant safety but also plant availability (economy). To overcome these sodium weak points, the appropriate countermeasure must be adopted to commercialized FBR plants. This report describes the working group activities for sodium/water reaction of steam generators (SG), in-service inspection for sodium components and sodium leak due to sodium components boundary failure. The prospect of each countermeasure is discussed in the viewpoint of the commercialized FBR plants. 1) Sodium/water reaction. The principle of the countermeasure for sodium/water reaction accidents was organized in the viewpoint of economy (the investment of SG and the plant availability). The countermeasures to restrain failure propagation were investigated for a large-sized SG. Preliminary analysis revealed the possibility of minimizing tubes failure propagation by improving the leak detection system and the blow down system. Detailed failure propagation analysis will be required and the early water leak detection system and rapid blow system must be evaluated to realize its performance. 2) In-service inspection (ISI and R). The viewpoint of the commercialized plant's ISI and R was organized by comparing with the prototype reactor's ISI and R method. We also investigated short-term ISI and R method without sodium draining to prevent the degrading of the plant availability, however, it is difficult to realize the with the present technology. Hereafter, the ISI and R of the commercialized plants must be defined by considering its characteristics. 3) Sodium leak from the components. This report organized the basic countermeasure policy for primary and secondary sodium leak accidents. Double-wall structure of sodium piping was

  19. Development of a standard data base for FBR core nuclear design. 9. Analysis of FCA XVII-1 experiments

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto; Oigawa, Hiroyuki; Iijima, Susumu

    1998-10-01

    Pnc had developed the adjusted nuclear cross-section library in which the results of the Jupiter experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of Fbr cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for FCA XVII-1 assembly. FCA XVII-1 is a representative mock-up of a MOX fuel sodium cooling FBR core. The criticality, reaction rate ratio, sodium void reactivity worth and 238 U Doppler reactivity worth of FCA XVII-1 were analyzed. The results of C/E values calculated by the standard analytical method for JUPITER experiments are similar to those calculated by the method of JAERI, except for the sodium void reactivity. So, further investigation for sodium void reactivity is necessary. Furthermore, sensitivity analysis shows the characteristics of FCA XVII-1 in comparison with ZPPR-9. (author)

  20. Multi-frequencies ECT algorithms to remove sodium noise in ISI of ferromagnetic SG tubes of FBR

    International Nuclear Information System (INIS)

    Mihalache, Ovidiu

    2012-01-01

    The paper presents developments and application of multi-frequency eddy current to be used during In-Service Inspection (ISI) of ferromagnetic steam generator (SG) tubes of Fast Breeder Reactors (FBR). Signal enhancement by means of multi-frequency ECT techniques are validated through 3D simulations of both signals and noise due to sodium forms around SG tube or SP. The purpose of such algorithms is to remove from ECT signal the electromagnetic noise resulting from sodium accumulated outside of SG tubes after SG vessel draining. Finite element method (FEM) simulations are used to analyse different sodium build-up scenarios observed experimentally, and to determine optimal multi-frequency ECT algorithms to suppress the most efficiently sodium noise. Also a new 'window multi-frequency' algorithm is applied and validated using 3-dimensional FEM simulations of SP and sodium forms. (author)

  1. Principles of MONJU maintenance. Characteristic of MONJU maintenance and reflection of LWR maintenance experience to FBR

    International Nuclear Information System (INIS)

    Nakai, Satoru; Nishio, Ryuichi; Uchihashi, Masaya; Kaneko, Yoshihisa; Yamashita, Hironobu; Yamaguchi, Atsunori; Aoki, Takayuki

    2014-01-01

    A sodium cooled fast breeder reactor (FBR) has unique systems and components and different degradation mechanism from light water reactor (LWR) so that need to establish maintenance technology in accordance with its features. The examination of the FBR maintenance technology is carried out in the special committee for considering the maintenance for Monju established in the Japan Society of Maintenology (JSM). As a result of the study such as extraction of Monju maintenance feature, maintenance technology benchmark between Monju and LWR components and survey of LWR maintenance experience, it is clear that principles of maintenance are same as LWR, necessity of LWR maintenance experience reflection and points to be considered in Monju maintenance. The road map to establish a FBR maintenance technology in the technical aspect became clear and it is vital to acquire operation and maintenance experience of the plant to implement this road map, and to establish a fast reactor maintenance. (author)

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

    International Nuclear Information System (INIS)

    Schubert, B.

    1988-01-01

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

  3. Degradation behavior of limestone concrete under limited time sodium exposure

    International Nuclear Information System (INIS)

    Das, S.K.; Sharma, A.K.; Ramesh, S.S.; Parida, F.C.; Kasinathan, N.; Chellapandi, P.

    2009-01-01

    Adequate safety measures are taken during design, fabrication, construction and operation of liquid sodium cooled fast breeder reactor (FBR). However, possibility of sodium leak from secondary heat transport circuits of FBR has not been completely ruled out. In the areas housing sodium pipelines such as Steam Generator Building (SGB), spilled liquid sodium not only reacts with air causing fire but also interacts with structural concrete resulting in its degradation. The structural concrete can be protected from sodium attack using sodium resistant sacrificial concrete layer or steel/refractory liners. Moreover, design and construction of sloping floor with sodium collection pit helps in minimizing the mass of sodium accumulated on the floor and exposure period. Sacrificial concrete layer on the structural concrete should meet key factors like economy, castability, easy removal of affected concrete in the event of a sodium fire and disposability of debris apart from its good resistance against hot burning sodium. Present study is directed towards testing of limestone concrete blocks (made out of 13% ordinary portland cement, 8% water, 48% coarse limestone and 31 % fine limestone aggregates)

  4. Cooling Performance of ALIP according to the Air or Sodium Cooling Type

    Energy Technology Data Exchange (ETDEWEB)

    Ye, Huee-Youl; Yoon, Jung; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    ALIP pumps the liquid sodium by Lorentz force produced by the interaction of induced current in the liquid metal and their associated magnetic field. Even though the efficiency of the ALIP is very low compared to conventional mechanical pumps, it is very useful due to the absence of moving parts, low noise and vibration level, simplicity of flow rate regulation and maintenance, and high temperature operation capability. Problems in utilization of ALIP concern a countermeasure for elevation of internal temperature of the coil due to joule heating and how to increase magnetic flux density of Na channel gap. The conventional ALIP usually used cooling methods by circulating the air or water. On the other hand, GE-Toshiba developed a double stator pump adopting the sodium-immersed self-cooled type, and it recovered the heat loss in sodium. Therefore, the station load factor of the plant could be reduced. In this study, the cooling performance with cooling types of ALIP is analyzed. We developed thermal analysis models to evaluate the cooling performance of air or sodium cooling type of ALIP. The cooling performance is analyzed for operating parameters and evaluated with cooling type. 1-D and 3-D thermal analysis model for IHTS ALIP was developed, and the cooling performance was analyzed for air or sodium cooling type. The cooling performance for air cooling type was better than sodium cooling type at higher air velocity than 0.2 m/s. Also, the air temperature of below 270 .deg. demonstrated the better cooling performance as compared to sodium.

  5. FBR Plant Engineering Center annual report 2012

    International Nuclear Information System (INIS)

    2013-12-01

    This annual report shows the last year's R and D activities of currently-reorganized FBR Plant Engineering Center, which was established on April 1, 2009. FBR Safety Technology Center was founded on April 1, 2013 by the consolidation of both the activities of 'former FBR Plant Engineering Center' and a portion of 'FBR Safety Evaluation Unit, Advanced Nuclear System Research and Development Directorate', especially concentrating on safety evaluations and analyses for severe accidents. As for FBR safety technology, it is necessary to continuously make an effort for compliance with new safety regulations in preparation for 'Monju' to restart, for safety enhancement evaluation and for safety technology upgrading. In this context, the new organization was founded in order to reinforce the safety evaluation capability, which will surely and steadily promote FBR safety-technology related activities. As a result, FBR Plant Engineering Center was abolished. This report summarizes the R and D activities at the former FBR Plant Engineering Center, aiming at contributing to the commercialization by using operation experiences and technology development results derived from the actual reactor 'Monju'. The activities are divided into five areas of operation-and-maintenance engineering, sodium engineering, reactor-core-and-fuel engineering, plant engineering, and safety engineering. This annual report is intended for a report of the activities of individual researcher in the center rather than that of the progress of the center as a whole. This will clarify the individual themes, progresses and problems of each researcher, which will, hopefully, facilitate communication with the outside researchers. (author)

  6. Fast reactor core design studies to cope with TRU fuel composition changes in the LWR-to-FBR transition period

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

    2009-01-01

    As part of the Fast Reactor Cycle Technology Development Project (FaCT Project), sodium-cooled fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. A 750 MWe mixed-oxide fuel core is firstly defined as a FaCT medium-size reference core and its neutronics characteristics are determined. The core is a high internal conversion type and has an average burnup of 150 GWD/T. The reference TRU fuel composition is assumed to come from the FBR equilibrium state. Compared to the LWR-to-FBR transition period, the TRU fuels in the FBR equilibrium period are multi-recycled through fast reactors and have a different composition. An available TRU fuel composition is determined by fast reactor spent fuel multi-recycling scenarios. Then the FaCT core corresponding to the TRU fuel with different compositions is set according to the TRU fuel composition changes in LWR-to-FBR transition period, and the key core neutronics characteristics are assessed. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters. (author)

  7. Sodium leakage experience at the prototype FBR Monju

    International Nuclear Information System (INIS)

    Miyakawa, A.; Maeda, H.; Kani, Y.; Ito, K.

    2000-01-01

    Monju is Japan's prototype fast breeder reactor: 280 MWe (714 MWt), fueled with mixed oxides of plutonium and uranium, cooled by liquid sodium. Construction was started in 1985 and initial criticality was attained in April 1994. On 8th December 1995, sodium leakage from a secondary circuit occurred in a piping room of the reactor auxiliary building. The secondary sodium leaked through a temperature sensor, due to the breakaway of the tip of the thermocouple well tube installed near the secondary circuit outlet of the intermediate heat exchanger (IHX). The reactor remained cooled and thus, from the viewpoint of radiological hazards, the safety of the reactor was secured. There was no release of radioactive material. There were no adverse effects for personnel and the surrounding environment. The thermocouple well tube failure resulted from high cycle fatigue due to flow induced vibration. It was found that this flow induced vibration was not caused by well-known Von Karman vortex shedding, but a symmetric vortex shedding. The design of the thermocouple well, which was subject to avoid this phenomenon, was reviewed. A new design guide against the flow-induced vibration was prepared by JNC (Japan Nuclear Cycle Development Institute). This is more comprehensive and definitive than the existing guide 'ASME N-1300' (Flow-induced vibration of tube and tube banks). New thermocouple well designs were proposed consistent with this design guide. To prevent a recurrence of the secondary sodium leakage incident, comprehensive design review activities were started for the purpose of checking the safety and reliability of the plant. As a result, several aspects to be improved were identified and improvements and countermeasures have been studied. The main improvements and countermeasures are as follows: To enable the operators to understand and react to incidents quickly, new sodium leakage detectors (TV monitors, smoke sensors) and a new surveillance system will be installed; To

  8. Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor. Comparisons of the decay heat removal characteristics on lead, lead-bismuth and sodium cooled reactors

    International Nuclear Information System (INIS)

    Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

    2000-04-01

    The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. In this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube failure accidents in a steam generator. In this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in Equivalent plant' with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. In conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to confirm the heat transfer reduction by the oxidized film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance. (author)

  9. FBR and RBR particle bed space reactors

    International Nuclear Information System (INIS)

    Powell, J.R.; Botts, T.E.

    1983-01-01

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10 0 K), high coolant-outlet temperatures (1500 to 3000 0 K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H 2 -cooled mode. The RBR will operate only in the open-cycle H 2 -cooled mode

  10. Status of conceptual safety design study of Japanese sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Kurisaka, Kenichi; Niwa, Hajime; Shimakawa, Yoshio

    2005-01-01

    In this paper, the current conceptual safety design and related evaluation of Japanese Sodium-cooled Fast Reactor which is studied in the framework of the Feasibility Study (FS) on commercialized Fast Reactor Cycle Systems in Japan are described. The purpose of the safety design is to establish a feasible safety concept of FBR which aims at a sustainable energy source of the next generations. The safety targets and the safety design principle are set aiming at realizing worldwide acceptability of the safety level. The basic safety design concept, which can meet the safety targets, was formulated taking along with the defense-in-depth philosophy as the basic safety design principle. In order to cope with wide range of energy and resource demands, there are some various designs both of oxide and metal fuel for JSFR. Some analytical results of typical design basis events, design extension conditions and core damage frequency estimation show the feasibility of the safety design concept for them. (author)

  11. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  12. Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

    International Nuclear Information System (INIS)

    Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

    2004-03-01

    The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

  13. Description of JNC's analytical method and its performance for FBR cores

    International Nuclear Information System (INIS)

    Ishikawa, M.

    2000-01-01

    The description of JNC's analytical method and its performance for FBR cores includes: an outline of JNC's Analytical System Compared with ERANOS; a standard data base for FBR Nuclear Design in JNC; JUPITER Critical Experiment; details of Analytical Method and Its Effects on JUPITER; performance of JNC Analytical System (effective multiplication factor k eff , control rod worth, and sodium void reactivity); design accuracy of a 600 MWe-class FBR Core. JNC developed a consistent analytical system for FBR core evaluation, based on JENDL library, f-table method, and three dimensional diffusion/transport theory, which includes comprehensive sensitivity tools to improve the prediction accuracy of core parameters. JNC system was verified by analysis of JUPITER critical experiment, and other facilities. Its performance can be judged quite satisfactory for FBR-core design work, though there is room for further improvement, such as more detailed treatment of cross-section resonance regions

  14. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  15. Development of flaw assesment methodology for elevated temperature components of FBR plants

    International Nuclear Information System (INIS)

    Shimakawa, Takashi; Takahashi, Yukio; Miura, Naoki; Nakayama, Yasunari; Sawai, Tatsuaki; Tooya, Yuuji

    1999-01-01

    Fracture mechanics is applicable for the safety assessment of FBR component if a crack is assumed to exist. Inelastic response should be taken into account due to high temperature operation of FBR components. However, methodology for the application of inelastic fracture mechanics has not been established sufficiently. CRIEPI has been conducted research projects to develop a flaw assessment guideline for FBR components. This guideline consists of evaluation methods for creep-fatigue crack propagation, ductile fracture and sodium leak rate. The summary of evaluation methods on creep-fatigue crack and ductile fracture is presented in this paper. (author)

  16. Raman distributed sensor system for temperature monitoring and leak detection in sodium circuits of FBR

    Energy Technology Data Exchange (ETDEWEB)

    Pandian, C.; Kasinathan, M.; Sosamma, S.; Babu Rao, C.; Jayakumar, T.; Murali, N.; Paunikar, V.; Kumar, S.; Rajan, K. K.; Raj, B. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2009-07-01

    Leak detection in coolant loops of nuclear reactors is critical for the safety and performance of the reactors. The feasibility of using Raman distributed temperature sensor (RDTS) has been studied on a 30 m test loop. Temperature in sodium circuits of fast Breeder Reactor (FBR) exceeds 550 C degrees, gold coated fiber is chosen as sensor fibers. Leak is simulated through an artificial micro fissure integrated in the test loop with provision for controlled leak rate. The results are discussed in the paper. The temperature response of RDTS is compared to the conventional thermocouple and their performance was found comparable. The feasibility of detecting the temperature differential of a controlled leak with RDTS is demonstrated

  17. Development of dose rate estimation system for FBR maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Iizawa, Katsuyuki [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Takeuchi, Jun; Yoshikawa, Satoru [Hitachi Engineering Company, Ltd., Hitachi, Ibaraki (Japan); Urushihara, Hiroshi [Ibaraki Hitachi Information Service Co., Ltd., Omika, Ibaraki (Japan)

    2001-09-01

    During maintenance activities on the primary sodium cooling system by an FBR Personnel radiation exposure arises mainly from the presence of radioactive corrosion products (CP). A CP behavior analysis code, PSYCHE, and a radiation shielding calculation code, QAD-CG, have been developed and applied to investigate the possible reduction of radiation exposure of workers. In order to make these evaluation methods more accessible to plant engineers, the user interface of the codes has been improved and an integrated system, including visualization of the calculated gamma-ray radiation dose-rate map, has been developed. The system has been verified by evaluating the distribution of the radiation dose-rate within the Monju primary heat transport system cells from the estimated saturated CP deposition and distribution which would be present following about 20 cycles of full power operation. (author)

  18. Development of dose rate estimation system for FBR maintenance

    International Nuclear Information System (INIS)

    Iizawa, Katsuyuki; Takeuchi, Jun; Yoshikawa, Satoru; Urushihara, Hiroshi

    2001-01-01

    During maintenance activities on the primary sodium cooling system by an FBR Personnel radiation exposure arises mainly from the presence of radioactive corrosion products (CP). A CP behavior analysis code, PSYCHE, and a radiation shielding calculation code, QAD-CG, have been developed and applied to investigate the possible reduction of radiation exposure of workers. In order to make these evaluation methods more accessible to plant engineers, the user interface of the codes has been improved and an integrated system, including visualization of the calculated gamma-ray radiation dose-rate map, has been developed. The system has been verified by evaluating the distribution of the radiation dose-rate within the Monju primary heat transport system cells from the estimated saturated CP deposition and distribution which would be present following about 20 cycles of full power operation. (author)

  19. Integrated thermal analysis of top-shield and reactor vault of Indian FBR-600

    International Nuclear Information System (INIS)

    Rajendrakumar, M.; Velusamy, K.; Selvaraj, P.

    2015-01-01

    The design for next generation fast breeder reactors (FBR-600) has been commenced with enhanced safety and improved economy as the main targets. The Top Shield (TS) of Prototype Fast Breeder Reactor (PFBR) is a box type structure consisting of Roof Slab (RS), Small Rotatable Plug (SRP), and Large Rotatable Plug (LRP). The large box type structure with many penetrations posed difficulties during manufacturing. Because of the required high load carrying capabilities, a dome shaped thick plate roof slab is conceived for FBR-600. Main Vessel (MV) which holds the primary sodium and associated components is welded to the RS through a triple joint. Reactor vault (RV) is a thick concrete structure which supports MV and Safety Vessel (SV). The temperature of RV concrete has to be less than 338 K (65°C) under normal operating heat loads (full and part load conditions) and less than 363 K (90°C) under Safety Grade Decay Heat Removal (SGDHR) conditions with one cooling loop in service. The temperature in the component penetrations of the RS should be greater than 120°C to avoid sodium aerosol deposition. Similarly, the temperature of the LRP and SRP has to be ∼120°C to protect the elastomeric seals provided to these structures. Further, the heat load to RV transferred by direct conduction by roof slab support has to be minimum. To meet these conflicting thermal requirements, detailed multi-physics CFD calculations have been performed to finalize, (i) the insulation requirements on the top of roof slab, (ii) number and position of reflective insulation plates below the bottom plate of roof slab/rotating plugs, (iii) air flow rate for various zones of the top shield and (iv) water flow rate and pitch of water cooling pipes for the reactor vault. (author)

  20. Development of blow down and sodium-water reaction jet analysis codes-Validation by sodium-water reaction tests (SWAT-1R)

    International Nuclear Information System (INIS)

    Hiroshi Seino; Akikazu Kurihara; Isao Ono; Koji Jitsu

    2005-01-01

    Blow down analysis code (LEAP-BLOW) and sodium-water reaction jet analysis code (LEAP-JET) have been developed in order to improve the evaluation method on sodium-water reaction event in the steam generator (SG) of a sodium cooled fast breeder reactor (FBR). The validation analyses by these two codes were carried out using the data of Sodium-Water Reaction Test (SWAT-1R). The following main results have been obtained through this validation: (1) The calculational results by LEAP-BLOW such as internal pressure and water flow rate show good agreement with the results of the SWAT- 1R test. (2) The LEAP-JET code can qualitatively simulate the behavior of sodium-water reaction. However, it is found that the code has tendency to overestimate the maximum temperature of the reaction jet. (authors)

  1. Present state of development of demonstration FBR and prospect of practical use

    International Nuclear Information System (INIS)

    Inagaki, Tatsutoshi

    1996-01-01

    As for the FBR development in Japan, the Atomic Energy Commission revised the long term plan on the research, development and utilization of atomic energy in June, 1994, and under the basic policy that through the considerable period of using LWRs together, FBRs will be adopted as the main nuclear power plants in future, it was decided to establish FBR technology system so that the practical use of FBRs becomes feasible by about 2030 through two demonstration FBRs following the experimental FBR 'Joyo' and the prototype FBR 'Monju'. The Monju started power generation and transmission in August, 1995, but secondary sodium leak accident occurred in December, 1995, and at present it is stopped. The demonstration FBR No. 1 is a top entry type loop reactor, and the power output is about 660 MWe. The start of construction is scheduled at the beginning of 2000s. The research on the whole plant design is carried out as the research on the optimization of demonstration FBR plant for three years from fiscal year 1994. The design of the demonstration FBR No. 1, the research and development for it, the prospect of the practical use and the research and development for the practical use are reported. (K.I.)

  2. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  3. A resting bottom sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Costes, D.

    2012-01-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  4. Training report of the FBR cycle training facility in 2004FY

    International Nuclear Information System (INIS)

    Watanabe, Toshio; Sasaki, Kazuichi; Sawada, Makoto; Ohtsuka, Jirou

    2004-07-01

    The FBR cycle training facility consists of sodium handling training facility and maintenance training facility, and is being contributed to train for the operators and maintenance workers of the prototype fast breeder reactor 'Monju'. So far, some training courses have been added to the both training courses of sodium handling technologies maintenance technologies in every year in order to carry out be significant training for preparation of Monju restarting. As encouragement of the sodium handling technology training in 2003FY, the sodium heat transfer basic course was equipped as the 9th sodium handling training course with the aims of learning basic principal technology regarding sodium heat transfer. While, for the maintenance training course, a named 'Monju Systems Learning Training Course', which aims to learn necessary knowledge as the engineers related Monju development, was provided newly in this year as an improvement concerned the maintenance course. In 2003FY, nine sodium handling technology training courses were carried out total 33 times and 235 trainees took part in those training courses. Also, nine training courses concerning the maintenance technology held 15 times and total 113 trainees participated. On the other hand, the 4th special lecture related sodium technology by France sodium school instructor was held on Mar. 15-17 and 34 trainees participated. Consequently, a cumulative trainees since October in 2000 opened the FBR cycle training facility reached to 1,236 so far. (author)

  5. Sodium test of the Super-Phenix full size primary pump shaft on the CPV-1 test rig at ENEA-Brasimone

    International Nuclear Information System (INIS)

    Contardi, T.; Rapezzi, L.; Partiti, C.; Zola, M.; Denimal, P.

    1984-01-01

    Tests on FBR Superphenix primary pump shaft were performed within the sodium-cooled FBR common research and development programs provided for by the cooperation agreement between ENEA and CEA. These tests were performed in CPV-1 plant ENEA - Brasimone Energy Research Center. The CPV-1 rig was built by FIAT-TTG and reproduces the reactor operating conditions (sodium-temperature and level, shaft inclination, etc..). Furthermore, CPV-1 rig's most interesting feature is its possibility to apply seismic stresses to test section by means of an oleodynamic actuator. Pivoterie-1 test section was made by JEUMONT-SCHNEIDER which built Superphenix pumps too; it was given to ENEA by FIAT-TTG. Seismic tests were performed with the cooperation of ISMES and FIAT-TTG. (author)

  6. Sodium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hammers, H.W.

    1982-01-01

    The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

  7. FBR/VHTR deployment scenarios in Japan

    International Nuclear Information System (INIS)

    Richards, Matt; Kunitomi, Kazuhiko

    2008-01-01

    Co-deployment of Fast Breeder Reactors (FBRs) and Very High Temperature Reactors (VHTRs) can be used as the nuclear technologies to meet a significant portion of Japan's future energy demands. The FBR provides the fissile fuel for energy security and sustainability, and can be used to provide a significant portion of the electricity demand. The VHTR can provide flexible energy outputs (electricity, hydrogen, and high-temperature heat) with high efficiency, can operate with a wide variety of fuel cycles, and can be sited at locations that have limited availability of cooling water. These features, combined with its passive safety and high degree of proliferation resistance, make the VHTR an ideal complement for co-deployment with the FBR in Japan and also a very low-risk technology of export to foreign countries. In addition to hydrogen production, the high-temperature thermal energy produced by the VHTR fleet can be used for a wide variety of process-heat applications, and the VHTR can play a key role for significantly reducing greenhouse-gas emissions. This paper describes assessments for deploying FBRs and VHTRs in Japan using a closed fuel cycle, with the FBRs supplying the fissile material to sustain the combined FBR/VHTR fleet. (author)

  8. Sodium pool fire analysis of sodium-cooled fast reactor by calculation

    International Nuclear Information System (INIS)

    Yu Hong; Xu Mi; Jin Degui

    2002-01-01

    Theoretical models were established according to the characteristic of sodium pool fire, and the SPOOL code was created independently. Some transient processes in sodium pool fire were modeled, including chemical reaction of sodium and oxygen; sodium combustion heat transfer modes in several kids of media; production, deposition and discharge of sodium aerosol; mass and energy exchange between different media in different ventilating conditions. The important characteristic parameters were calculated, such as pressure and temperature of gas, temperature of building materials, mass concentration of sodium aerosol, and so on. The SPOOL code, which provided available safety analysis tool for sodium pool fire accidents in sodium-cooled fast reactor, was well demonstrated with experimental data

  9. Investigation/evaluation of water cooled fast reactor in the feasibility study on commercialized fast reactor cycle systems. Intermediate evaluation of phase-II study

    International Nuclear Information System (INIS)

    Kotake, Syoji; Nishikawa, Akira

    2005-01-01

    Feasibility study on commercialized fast reactor cycle systems aims at investigation and evaluation of FBR design requirement's attainability, operation and maintenance, and technical feasibility of the candidate system. Development targets are 1) ensuring safety, 2) economic competitiveness, 3) efficient utilization of resources, 4) reduction of environmental load and 5) enhancement of nuclear non-proliferation. Based on the selection of the promising concepts in the first phase, conceptual design for the plant system has proceeded with the following plant system: a) sodium cooled reactors at large size and medium size module reactors, b) a lead-bismuth cooled medium size reactor, c) a helium gas cooled large size reactor and d) a BWR type large size FBR. Technical development and feasibility has been assessed and the study considers the need of respective key technology development for the confirmation of the feasibility study. (T. Tanaka)

  10. Shape optimization of a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Schmitt, D.; Allaire, G.; Pantz, O.; Pozin, N.

    2013-01-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth. Usual optimization methods for core conception are based on a parametric description of a given core design. New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints. First studies show that these methods could be applied to sodium cooled core conception. In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a given realistic core layout. Its characteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas. (authors)

  11. Investigation of safety measures to severe accident of Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    So as to plan the accident management to severe accident of Fast Breeder Reactor (FBR), it is primary important to understand the progression of severe accident (SA) precisely. In this study, it has been aimed to reveal two items that work as keys in the evaluation of SA in sodium cooled FBR. One is the cool-ability of degraded core on the core support plate by sodium natural circulation in the post accident heat removal (PAHR) phase. An obstacle that hinders the smooth heat transfer from fuel debris to coolant is the formation of sodium-uranate by chemical reaction between sodium and fuel. Following the measurement of physical values of sodium-uranate in FY 2011, experiments has been performed to reveal the conditions for sodium-uranate formation on fuel debris in sodium pool simulating the actual situation of the degraded core. The cool-ability of the debris bed was analyzed using the Lipinski 1-D model. Another research performed in this study is the measurement of fission product (cesium and antimony) evaporation rates from FBR fuel as a function of temperature, because presently the fission product evaporation rates data for LWR is also temporarily used for FBR SA analysis. The measurement was performed using the irradiated fuels in the Test Reactor JOYO. (author)

  12. Parametric study of sodium aerosols in the cover-gas space of sodium-cooled reactors

    International Nuclear Information System (INIS)

    Sheth, A.

    1975-03-01

    A mathematical model has been developed to describe the behavior of sodium aerosols in the cover-gas space of a sodium-cooled reactor. A review of the literature was first made to examine methods of aerosol generation, mathematical expressions representing aerosol behavior, and pertinent experimental investigations of sodium aerosols. In the development of the model, some terms were derived from basic principles and other terms were estimated from available correlations. The model was simulated on a computer, and important parameters were studied to determine their effects on the overall behavior of sodium aerosols. The parameters studied were sodium pool temperature, source and initial size of particles, film thickness at the sodium pool/cover gas interface, wall plating parameters, cover-gas flow rate, and type of cover gas (argon and helium). The model satisfactorily describes the behavior of sodium aerosol in argon, but not in helium. Possible reasons are given for the failure of the model with helium, and further experimental work is recommended. The mathematical model, with appropriate modifications to describe the behavior of sodium aerosols in helium, would be very useful in designing traps to remove aerosols from the cover gas of sodium-cooled reactors. (U.S.)

  13. Design study on sodium-cooled middle-scale modular reactor

    International Nuclear Information System (INIS)

    Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

    2002-05-01

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled middle-scale modular reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2001, which is the first year of Phase 2. As the construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in Phase 1, was about 10% higher than that of the sodium-cooled large-scale reactor, a new concept of the middle-scale modular reactor, which is expected to be equal to the large-scale reactor from a viewpoint of economic competitiveness, has been re-constructed based on the design of the advanced loop type reactor. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  14. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki [Nuclear Research Group, FMIPA, Bandung Institute of Technology Jl. Ganesha 10, Bandung 40132 (Indonesia); Miura, Ryosuke; Takaki, Naoyuki [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, H. [Emerritus Prof. of Research Laboratory for Nuclear Reactors, Tokyo Inst. of Technology (Japan)

    2015-09-30

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  15. Study on chemical reactivity control of liquid sodium. Research program

    International Nuclear Information System (INIS)

    Saito, Jun-ichi; Ara, Kuniaki; Sugiyama, Ken-ichiro; Kitagawa, Hiroshi; Oka, Nobuki; Yoshioka, Naoki

    2007-01-01

    Liquid sodium has the excellent properties as coolant of the fast breeder reactor (FBR). On the other hand, it reacts high with water and oxygen. So an innovative technology to suppress the reactivity is desired. The purpose of this study is to control the chemical reactivity of liquid sodium by dispersing the nanometer-size metallic particles (we call them Nano-particles) into liquid sodium. We focus on the atomic interaction between Nano-particles and sodium atoms. And we try to apply it to suppress the chemical reactivity of liquid sodium. Liquid sodium dispersing Nano-particles is named 'Nano-fluid'. Research programs of this study are the Nano-particles production, the evaluation of reactivity suppression of liquid sodium and the feasibility study to FBR plant. In this paper, the research programs and status are described. The important factors for particle production were understood. In order to evaluate the chemical reactivity of Nano-fluid the research programs were planned. The feasibility of the application of Nano-fluid to the coolant of FBR plant was evaluated preliminarily from the viewpoint of design and operation. (author)

  16. Base technology development of new materials for FBR performance innovations

    International Nuclear Information System (INIS)

    Kano, Shigeki; Koyama, Masahiro; Nomura, Shigeo; Morikawa, Satoru; Ueno, Fumiyoshi

    1989-01-01

    This paper describes the base technology development of new materials for FBR performance innovations at the Power Reactor and Nuclear Fuel Development Corporation. The contents are as follows: (1) development of sodium and radiation resistant new materials, (2) development of high performance shielding material, (3) development of high performance control material, (4) development of new functional materials for reactor instrumentation. (author)

  17. Simplified numerical simulation of hot channel in sodium cooled reactor

    International Nuclear Information System (INIS)

    Fonseca, F. de A.S. da; Silva Filho, E.

    1988-12-01

    The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt

  18. Under-Sodium-Viewing as one technique for periodic inspections in sodium-cooled fast reactors-- possibilities and limits

    International Nuclear Information System (INIS)

    Weiss, H.

    1979-07-01

    Periodic inspections are gaining increasingly technical importance for fast sodium cooled reactors. Among others the reactor tank and its internals have to be inspected, whereby licensing experts partly are requesting the standards of Light Water Reactors. This leads to difficulties in sodium cooled reactors because of the non-transparent coolant sodium and their compact structure. In order to avoid the complete dumping of the sodium, the under sodium viewing shall be applied besides other inspection methods. Since this is a new method, which is still in its development phase, this report presents and discusses the technical and physical basis and outlines possibilities and limits [de

  19. Evolution of design of steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.; Vaidyanathan

    1997-01-01

    The first sodium cooled reactor was the experimental breeder reactor (EBR-I) in usa which was commissioned in 1951 and was incidentally the first nuclear reactor to generate electrical energy. This was followed by fast breeder reactors in USSR, UK, france, USA, japan, germany and India. The use of sodium as a coolant is due to its low moderation which helps in breeding fissile fuel from fertile materials and also its high heat transfer coefficient at comparatively low velocities. The good heat transfer properties introduce thermal stresses when there are rapid changes in the sodium temperatures. Also sodium has a chemical affinity with air and water. The steam generators for sodium cooled reactors have to allow for these novel conditions and in addition, unlike other components. Choices have to be made whether it is a recirculation type as in most fossil plants or an once through unit, the power rating, shape of the tube (straight, helical, U-tube), materials (Ferritic or austenitic), with free level of sodium or not, sodium on tube side or shell side and so on. With higher pressures and steam temperatures reheating steam after partial expansion in the turbine becomes essential as in conventional turbines. For this purpose the choice of reheating fluid viz sodium or live main steam has to be made. This paper traces the evolution of steam generator designs in the different sodium cooled reactors (chronologically) and the operation experience. 16 figs., 1 tab

  20. Structural dynamics in FBR

    International Nuclear Information System (INIS)

    Bhoje, S.B.

    2003-01-01

    In view of thin walled large diameter shell structures with associated fluid effects, structural dynamics problems are very critical in a fast breeder reactor. Structural characteristics and consequent structural dynamics problems in typical pool type Fast Breeder Reactor are highlighted. A few important structural dynamics problems are pump induced as well as flow induced vibrations, seismic excitations, pressure transients in the intermediate heat exchangers and pipings due to a large sodium water reaction in the steam generator, and core disruptive accident loadings. The vibration problems which call for identification of excitation forces, formulation of special governing equations and detailed analysis with fluid structure interaction and sloshing effects, particularly for the components such as PSP, inner vessel, CP, CSRDM and TB are elaborated. Seismic design issues are presented in a comprehensive way. Other transient loadings which are specific to FBR, resulting from sodium-water reaction and core disruptive accident are highlighted. A few important results of theoretical as well as experimental works carried out for 500 MWe Prototype Fast Breeder Reactor (PFBR), in the domain of structural dynamics are presented. (author)

  1. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  2. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

    The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

  3. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

    Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

  4. Correlations among FBR core characteristics for various fuel compositions

    International Nuclear Information System (INIS)

    Maruyama, Shuhei; Ohki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu

    2012-01-01

    In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations. (author)

  5. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  6. Design study on sodium-cooled large-scale reactor

    International Nuclear Information System (INIS)

    Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

    2002-05-01

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2001, which is the first year of Phase 2. In the JFY2001 design study, a plant concept has been constructed based on the design of the advanced loop type reactor, and fundamental specifications of main systems and components have been set. Furthermore, critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  7. Development of evaluation methodology to assess the sodium fire suppression performance of leak collection tray

    International Nuclear Information System (INIS)

    Parida, F.C.; Rao, P.M.; Ramesh, S.S.; Somayajulu, P.A.; Malarvizhi, B.; Kannan, S.E.

    2005-01-01

    Full text of publication follows: Leakage of hot liquid sodium and its subsequent combustion in the form of a pool cannot be completely ruled out in a Fast breeder Reactor (FBR) plant in spite of provision for adequate safety measures. To protect the plant system from the hazardous effects of flame, heat and smoke, one of the passive protection devices used in FBR plants is the Leak Collection Tray (LCT). The design of LCT is based on immediate channeling of burning liquid sodium on the funnel shaped sloping cover tray (SCT) to the bottom sodium hold-up vessel (SHV) in which self-extinction of the fire occurs due to oxygen starvation. The SCT has one or three drain pipes and air vent pipes depending on the type of design. In each experiment, a known amount ranging from 30 to 40 kg of hot liquid sodium at 550 deg. C was discharged on the LCT in the open air. Continuous on-line monitoring of temperature at strategic locations (∼ 28 points) was carried out. Colour video-graphy was employed for taking motion pictures of various time-dependent events like sodium dumping, appearance of flame and release of smoke through vent pipes. After self-extinction of sodium fire, the LCT was allowed to cool overnight in an argon atmosphere. Solid samples of sodium debris in the SCT and SHV were collected by manual core drilling machine. The samples were subjected to chemical analysis for determination of unburnt and burnt sodium. The sodium debris removed from SCT and SHV were separately weighed. To assess the performance of the LCT, two different geometrical configurations of SCT, one made up of stainless steel an the other of carbon steel, were used. Three broad phenomena are identified as the basis of evaluation methodology. These are (a) thermal transients, i.e. heating and cooling of the bulk sodium in SCT and SHV respectively, (b) post test sodium debris distribution between SCT and SHV as well as (c) sodium combustion and smoke release behaviour. Under each category

  8. Development of evaluation methodology to assess the sodium fire suppression performance of leak collection tray

    Energy Technology Data Exchange (ETDEWEB)

    Parida, F.C.; Rao, P.M.; Ramesh, S.S.; Somayajulu, P.A.; Malarvizhi, B.; Kannan, S.E. [Engineering Safety Division, Safety Group, Indira Gandhi Centre for Atomic Research, Kalpakkam - 603102, Tamilnadu (India)

    2005-07-01

    Full text of publication follows: Leakage of hot liquid sodium and its subsequent combustion in the form of a pool cannot be completely ruled out in a Fast breeder Reactor (FBR) plant in spite of provision for adequate safety measures. To protect the plant system from the hazardous effects of flame, heat and smoke, one of the passive protection devices used in FBR plants is the Leak Collection Tray (LCT). The design of LCT is based on immediate channeling of burning liquid sodium on the funnel shaped sloping cover tray (SCT) to the bottom sodium hold-up vessel (SHV) in which self-extinction of the fire occurs due to oxygen starvation. The SCT has one or three drain pipes and air vent pipes depending on the type of design. In each experiment, a known amount ranging from 30 to 40 kg of hot liquid sodium at 550 deg. C was discharged on the LCT in the open air. Continuous on-line monitoring of temperature at strategic locations ({approx} 28 points) was carried out. Colour video-graphy was employed for taking motion pictures of various time-dependent events like sodium dumping, appearance of flame and release of smoke through vent pipes. After self-extinction of sodium fire, the LCT was allowed to cool overnight in an argon atmosphere. Solid samples of sodium debris in the SCT and SHV were collected by manual core drilling machine. The samples were subjected to chemical analysis for determination of unburnt and burnt sodium. The sodium debris removed from SCT and SHV were separately weighed. To assess the performance of the LCT, two different geometrical configurations of SCT, one made up of stainless steel an the other of carbon steel, were used. Three broad phenomena are identified as the basis of evaluation methodology. These are (a) thermal transients, i.e. heating and cooling of the bulk sodium in SCT and SHV respectively, (b) post test sodium debris distribution between SCT and SHV as well as (c) sodium combustion and smoke release behaviour. Under each category

  9. Evaluation of the commercial FBR introduction date

    International Nuclear Information System (INIS)

    White, M.K.; Merrill, E.T.

    1981-09-01

    This report examines one criterion for introducing a commercial FBR: economic competitiveness with a Light Water Reactor (LWR). For this analysis, the commercial FBR is assumed to be the fifth-of-a kind replicate which represents an economically mature plant. This FBR is deemed economically competitive when its life-cycle energy cost is less than or equal to that of an LWR. Results of this analysis are used in a comparative analysis of alternative FBR development stategies. The strategies evaluated in these studies assume both 1000- and 1457-MWe FBRs. Since the capital costs per kilowatt, and therefore the energy costs, for these two FBR sizes are different, they will become economically competitive at different times. The probability density function for the 1457-MW(e) FBR has an expected value date or weighted average date of 2030, compared with 2033 for the probability density function for the 1000-MW(e) FBR

  10. Transformation of sodium from the Rapsodie fast breeder reactor into sodium hydroxide

    International Nuclear Information System (INIS)

    Roger, J.; Latge, C.; Rodriguez, G.

    1994-01-01

    One of the major problems raised by decommissioning a fast breeder reactor (FBR) concerns the disposal of the sodium coolant. The Desora operation was undertaken to eliminate the Rapsodie primary sodium as part of the partial decommissioning program, and to develop an operational sodium treatment unit for other needs. The process involves reacting small quantities of sodium in water inside a closed vessel, producing aqueous sodium hydroxide and hydrogen gas. It is described in this work. (O.L.). 4 figs

  11. Development of the tool for generating ORIGEN2 library based on JENDL-3.2 for FBR

    International Nuclear Information System (INIS)

    Ohkawachi, Yasushi; Fukushima, Manabu

    1999-05-01

    ORIGEN2 is one of the most widely-used burnup analysis code in the world. This code has one-grouped cross section libraries compiled for various types of reactors. However, these libraries have some problems. One is that these libraries were developed from old nuclear data libraries (ENDF/B-IV,V) and the other is that core and fuel designs from which these libraries are generated do not match the current analysis. In order to solve the problems, analysis tool is developed for generating ORIGEN2 library from JENDL-3.2 considering multi-energy neutron spectrum. And eight new libraries are prepared using this tool for analysis of sodium-cooled FBR. These new libraries are prepared for eight kinds of cores in total. Seven of them are made by changing core size (small core - large core), fuel type (oxide, nitride, metal) and Pu vector as a parameter. The eighth one is a Pu burner core. Burnup calculation using both new and original libraries, shows large difference in buildup or depletion numbers of nuclides among the libraries. It is estimated that the analysis result is greatly influenced by the neutron spectrum which is used in collapse of cross section. By using this tool or new libraries, it seems to improve evaluation accuracy of buildup or depletion numbers of nuclides in transmutation research on FBR fuel cycle. (author)

  12. FBR metallic materials test manual (English version)

    International Nuclear Information System (INIS)

    Odaka, Susumu; Kato, Shoichi; Yoshida, Eiichi

    2003-06-01

    For the development of the fast breeder reactor, this manual describes the method of in-air and in-sodium material tests and the method of organization the data. This previous manual has revised in accordance with the revision of Japanese Industrial Standard (JIS) and the conversion to the international unit. The test methods of domestic committees such as the VAMAS (Versailles Project on Advanced Materials and Standards) workshop were also refereed. The material test technologies accumulated in this group until now were also incorporated. This English version was prepared in order to provide more engineers with the FBR metallic materials test manual. (author)

  13. Countermeasures to cope with issues for the FBR cycle system and transitional core during FBR introductory period

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sasahira, Akira; Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi

    2007-01-01

    The introduction of fast breeder reactors (FBRs) requires Pu be recovered from light water reactors (LWRs) spent fuel. The 'Flexible Fuel Cycle Initiative (FFCI)' can supply enough Pu and holds no surplus Pu, while responding flexibly to future technical and social uncertainties. In this paper, the potential of FFCI to increase economy of the fuel cycle system was investigated. On the other hand, during the FBR introductory period, Pu from LWR spent fuel is used for startup of FBRs. But the FBR core being loaded with Pu from LWR spent fuel has a larger burnup reactivity than the core being loaded with Pu from the FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of the FBR core. In this paper, an FBR transitional core concept to handle the issues of the FBR introductory period was investigated. The results obtained through this study are as follows. (1) The FFCI has a potential to flatten the reprocessing amount of LWR spent fuel and to increase economy of the next fuel cycle system. (2) Minor actinides - mixed FBR transitional core has a potential to maintain the operation cycle length even supposing use of Pu from LWR spent fuel. (author)

  14. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

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

    2003-05-01

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

  15. Status of feasibility study for various technical options of FBR systems

    International Nuclear Information System (INIS)

    Kani, Yoshio

    2000-01-01

    JNC (Japan Nuclear Cycle Development Institute) has started a new research project of feasibility studies (F/S) for a wide variety option of fast breeder reactor (FBR) and related fuel cycle in order to develop an economically competitive FBR cycle system fro commercialization. JNC and the electric untilities in Japan have established a new organization in JNC to perform the F/S since July 1, 1999. The organization has undertaken feasibility studies (F/S) in order to determine promising FBR cycle concepts and define necessary RandD tasks. The long-term targets of commercialized FBR cycle system are set as ensuring safety, economic competitiveness relative to future LWRs, efficient utilization of resources, reduction in environmental burden, and enhancement of nuclear non-proliferation. This paper describes the progress of design studies for a wide variety of technical options of FBR plants in the framework of the F/S. We make efforts towards considering all key issues so as not to fail to notice the best concept in a commercialized stage. In the study of technical options, the identified coolant types are sodium, heavy metal (lead and lead-bismuth), gas (carbon dioxide and helium ) and water (boiling water, pressurized water and supercritical water). The classified types of fuel are mixed oxide, nitride and metal. Design studies of small size modular plant concepts are also performed. We study many reactor concepts in combination with a coolant type and a fuel type, understand characteristics of each reactor concept based on our experience and an extensive survey of literature, and make a draft design of each reactor concept for rough estimation of construction costs. We also check how far the concept accomplishes each index (safety, economy, resource utilization, etc.) of design requirements, and will select several promising reactor concepts. (author)

  16. Mapping of sodium void worth and doppler effect for sodium-cooled fast reactor - 15458

    International Nuclear Information System (INIS)

    Krepel, J.; Pelloni, S.; Bortot, S.; Panadero, A.L.; Mikityuk, K.

    2015-01-01

    The sodium-cooled fast reactor (SFR) represents the reference and the most technologically mastered system among the Generation-IV reactors. Nevertheless, the sodium void worth in the fuel regions of SFR is usually positive. To overcome this safety drawback, low-void sodium-cooled fast spectrum core (CFV) was proposed by CEA. Such a CFV core is used in the frame of WP6 'Core safety' of the FP7 Euratom ESNII+ project as a reference SFR design. The overall sodium void effect is negative for the CFV core. Nevertheless, locally it is positive in the fuel region and negative in the sodium plenum. Similarly, also the Doppler effect is spatially dependent and it varies between the inner and outer fuel regions and between the middle and lower blankets. Accordingly, knowledge of the local distributions or actually mappings of the two safety-related parameters will be necessary, before safety assessment and transient analysis can be done. In this study these maps have been produced using the deterministic code ERANOS. The obtained mapping shows strong local dependency of both safety-related effects. A sensitivity of the void effect to the sodium plenum modeling was also demonstrated. The results may serve as an input for the transient analysis of the CFV core or as a cross-check for the Monte Carlo method based maps. (authors)

  17. Analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1976-09-01

    The aim of this analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors is to get results on the possible extent of blockages and the time necessary for growth which may be used for a safety evaluation. After an introduction where the thermohydraulic and physical/chemical aspects of the problems are considered, the causes for the local cooling disturbances and the phenomena arising with it are freated in more detail. (orig./TK) [de

  18. Hydrogen detector for sodium cooled reactors

    International Nuclear Information System (INIS)

    Roy, P.; Rodgers, D.N.

    1975-01-01

    An improved hydrogen detector for use in sodium cooled reactors is described. The improved detector basically comprises a diffusion tube of either pure nickel or stainless steel having a coating on the vacuum side (inside) of a thin layer of refractory metal, e.g., tungsten or molybdenum. The refractory metal functions as a diffusion barrier in the path of hydrogen diffusing from the sodium on the outside of the detector into the vacuum on the inside, thus by adjusting the thickness of the coating, it is possible to control the rate of permeation of hydrogen through the tube, thereby providing a more stable detector. (U.S.)

  19. Design study on sodium-cooled middle-scale modular reactor

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki

    2003-09-01

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2002, which is the second year of Phase 2. The construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in JFY2002, was almost achieved the economical goal. But its achievability was not sufficient to accept the concept. In order to reduce the construction cost, the plant concept has been re-constructed based on the 50 MWe plant studied in JFY2002. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects have been examined and evaluated. In addition, in order to achieve the further cost reduction, the plant with simplified secondary system, the plant with electric magnetic pump in secondary system, and the fuel handling system are examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  20. Scenario study on the FBR deployment

    International Nuclear Information System (INIS)

    Ono, Kiyoshi; Kofuji, Hirohide; Otaki, Akira; Yonezawa, Shigeaki; Shinoda, Yoshihiko; Hirao, Kazunori; Ikegami, Tetsuo

    2000-12-01

    This study on success scenarios for the Fast Breeder Reactor (FBR) deployment was performed taking account of future situation of fossil, renewable and nuclear energies in Japan as well as the world from the viewpoints of the following four items; economics, environment, energy security and restriction of natural uranium resources. In the economics scenario, if carbon tax is added to generating cost of LNG, coal and oil and the economics of FBR cycle is competitive with LWR cycle in the future, FBR cycle will be expected to introduce as the middle and base load power plant. In the environment scenario, there is also any possibility that FBR cycle which can burn and transmute minor actinide and fission product elements will be introduced in order to reduce the burden of deposit facility and the toxicity of high-level waste. In the uranium resources restriction scenario, FBR cycle needs to be deployed at the latest in the middle of 21st century from the viewpoint of the restriction of natural uranium resources. This study was carried out in a part of JNC's feasibility study on commercialized FBR cycle system. (author)

  1. Material properties of oxide dispersion strengthened (ODS) ferritic steels for core materials of FBR. Tensile properties of sodium exposed and nickel diffused materials

    International Nuclear Information System (INIS)

    Kato, Shoichi; Yoshida, Eiichi

    2002-12-01

    An oxide dispersion strengthened (ODS) ferritic steel is candidate for a long-life core materials of future FBR, because of good swelling resistance and high creep strength. In this study, tensile tests were carried out the long-term extrapolation of sodium environmental effects on the mechanical properties of ODS steels. The tested heats of materials are M93, M11 and F95. The specimens were pre-exposed to sodium for 1,000 and 3,000 hours under non-stress conditions. The pre-exposure to sodium was conducted using a sodium test loop constituted by austenitic steels. For the conditions of sodium exposure test, the sodium temperature was 650 and 700degC, the oxygen concentration in sodium was about 1 ppm and sodium flow rate on the surface of specimen was less than 1x10 -4 m/seconds (nearly static). Further the specimen with the nickel diffused was prepared, which is simulate to nickel diffusing through sodium from the surface of structural stainless steels. The main results obtained were as follows; (1) The tensile strength and the fracture elongation after sodium exposure (maximum 3,000 hours) were same as that of as-received materials. If was considered that the sodium environmental effect is negligible under the condition of this study. (2) Tensile properties of nickel diffused specimens were slightly lower than that of the as-received specimens, but it remains equal to that of thermal aging specimens. (3) The change in microstructure such as a degraded layer was observed on the surface of nickel diffused specimen. In the region of the degraded layer, phase transformations from the α-phase to the γ-phase were recognized. But, the microscopic oxide particles were observed same as that of α-phase base metal. (author)

  2. Parameters' influence estimation on Puf supply and demand in transitional period from LWR to FBR in Japan

    International Nuclear Information System (INIS)

    Kobayashi, Hiroaki; Ohta, Hirokazu; Inoue, Tadashi

    2009-01-01

    Plutonium fissile (Puf) amounts to balance supply and demand during transition period were evaluated with different parameters. Estimated total Puf demand in transitional period was sensitive to deployment speed of FBR. Because FBRs will be deployed as replacements of old LWRs for keeping total capacity, deployment history of existing LWRs should be taken into consideration. According to the estimation, LWR fuel burnup and utilized capacity are not big issue. Because certain amount of LWR spent fuel will remain in early phase of transitional period, there is enough time for preparing Puf supply. On the other hand, FBR fuel cycle time (SF cooling time + fuel fabrication time) have large impact on Puf supply. Fuel cycle technologies including transportation for applying to short cooling spent fuels should be developed. (author)

  3. Study on thermalhydraulics of natural circulation decay heat removal in FBR. Experiment with water of typical reactor trip in the demonstration FBR

    International Nuclear Information System (INIS)

    Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

    2010-01-01

    Intending to enhance safety and to reduce costs, an FBR plant is being developed in Japan. In relies solely on natural circulation of the primary cooling loop to remove a decay heat of the core after reactor trips. A water test was carried out to advance the development. The test used a 1/10 reduced scale model simulating the core and cooling systems. The experiments simulated representative accidents from steady state to decay heat removal through reactor trip and clarified thermal-hydraulic issues on the thermal circulation performance. Some modifications of the system design were proposed for solving serious problems of natural circulation. An improved design complying with the suggestions will make it possible for natural circulation of the cooling systems to remove the decay heat of the core without causing and unstable or unpredictable change. (author)

  4. Development of bellows for sodium valves in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, S; Mukai, K; Fukada, T; Takahashi, T

    1980-02-01

    Sodium valves for FBR are required to isolate sodium side from the atmosphere completely throughout its lifetime because of preventing sodium leakage with or without radioactivity. A great number of sodium valves have been used in FBR test facilities at O-arai Engineering Center of PNC and many troubles have occurred through their operational experience. Most of the cause of the troubles are the bellows failure followed by sodium leakage. A research and development program on bellows was started to clarify many uncertain factors of its performance and to establish the feasibility of bellows used in sodium. In this program Small Bellows Test Loop was built to perform low cycle fatigue tests on bellows under high temperature conditions. In this report some examples of the investigation of failed bellows occurred at O-arai Engineering Center of PNC are described. The research and development program on bellows is also explained with the summary of recent test results. (author)

  5. Fiber optic sensors for monitoring sodium circuits and power grid cables

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, M.; Sosamma, S.; Pandian, C.; Vijayakumar, V.; Chandramouli, S.; Nashine, B. K.; Rao, C. B.; Murali, N.; Rajan, K. K.; Jayakumar, T. [IGCAR, Kalpakkam (India)

    2011-07-01

    At Kalpakkam, India, a programme on development of Raman Distributed Temperature sensor (RDTS) for Fast Breeder Reactors (FBR) application is undertaken. Leak detection in sodium circuits of FBR is critical for the safety and performance of the reactors. It is demonstrated that RDTS can be usefully employed in monitoring sodium circuits and in tracking the percolating sodium in case of any leak. Aluminum Conductor Steel Reinforced (ACSR) cable is commonly used as overhead power transmission cable in power grid. A second application demonstrates the suitability of using RDTS to monitor this transmission cable for any defect. (authors)

  6. Methods for the sodium cooled fast reactor fire safety provisions

    International Nuclear Information System (INIS)

    Gryaznov, B.V.; Dergachev, N.P.

    1983-01-01

    Problems of fire safety provision on NPPs with sodium cooled fast reactor are under discussion. Methods of sodium leak localization, measures eliminating sodium flaring up during leaks and main means of sodium fire extinguishing are considered. An extinguishing of sodium flaring up is performed by means of sodium temperatUre decrease and by limitation of hydrogen access to the flaring up surface. A conclusion is made that the most effective methods of extinguishing are the following: self-extinguishing (due to hydrogen burning out in a limiting volume); extinguishing by a gas mixture of nitrogen and carbonic acid (initial filling and blowing of rooms during sodium flaring up); extinguishing by special powders

  7. Experiments in LEENA facility with modified wire type leak detector layout in large sodium pipelines

    International Nuclear Information System (INIS)

    Vijayakumar, G.; Chandramouli, S.; Nashine, B.K.; Selvaraj, P.; Rajan, K.K.

    2017-01-01

    Highlights: • FBR large horizontal secondary pipeline were simulated and five sodium leak experiments were conducted by providing modified wire type leak detector layout at 550 °C. • Early detection of sodium leak is needed for minimizing the sodium leaked out and consequent damages. • PFBR leak detector layout on large horizontal pipelines can detect a leak rate of 200 g/h within 6 h. • By reducing the distance between leak point and detector to half, detection time was reduced to 1/6th and found that a leak rate of 200 g/h can be detected in one hour. • A relationship between leak rate and detection time was established based on experimental results. - Abstract: Sodium cooled Fast Breeder Reactors (SFRs) are envisaged in the second phase of Indian nuclear power programme. Liquid sodium is used as the coolant in the SFRs due to its favourable nuclear properties and excellent heat transfer properties. Leaks in sodium systems have the potential of being exceptionally hazardous due to the reaction of liquid sodium with oxygen and water vapour in the air. When a sodium leak occurs, the sodium leak rate, the total quantity of sodium leaked and leak detector layout governs the detection time. Other factors to be considered are insulation material packing condition, distance between the leak point and detector, heater layout, pipe geometry, temperature etc. Potential regions of leakage in Fast Breeder Reactor (FBR) sodium circuits are near welds, high stress areas and regions subjected to thermal striping. Early detection of leak is needed for minimizing the quantity of sodium leaked to outside and consequent damages. Three wire type leak detectors (WLDs positioned at 90°, 180° and 270°) working on conductivity principle are used for detecting sodium leak in the large horizontal secondary sodium pipelines of Prototype Fast Breeder Reactor (PFBR). It was found from the upper boundary curve based on LEENA (LEak Experiments in NAtrium) facility experimental

  8. Materials Options of Steam Generator for Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Fu Xiaogang; Long Bin; Han Liqing; Qin Bo; Zhang Jinquan; Wang Shuxing

    2013-01-01

    Overview of the material options of steam generator for sodium-cooled fast reactors, the method to calculate the service life, the thinning of wall thickness and the sodium corrosion rate, the degradation of mechanical properties (thermal aging and decarburization) and the calculation results of theoretical models

  9. Apparatus for removing impurities in the sodium of sodium cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yamauchi, A

    1970-11-11

    An apparatus is provided for removing oxygen from liquid sodium flowing in a sodium cooled reactor. The removal of oxygen is complete with high efficiency. The liquid sodium to be purified is disposed outside a cylindrical wall and negatively charged, whereas sodium as a reducing material is disposed inside the same wall. The cylindrical wall is made of zirconia-calcia (ZrO/sub 2/)sub(0.87)(CaO)sub(0.13) solid electrolyte, the cylinder having a thickness of 2.5mm, a diameter of 3cm and a depth of 20cm under the sodium level. Electric resistance of the solid electrolyte is 2.3 ohm at 500/sup 0/C. A current of 1A by the application of 25 volts treats 0.3g of oxygen. Consequently, 1 liter or 1kg of liquid sodium containing 1,000ppm of oxygen can be purified for about 3 hours at an electrical consumption of 7.5 watt-hour. In one embodiment, a cylindrical electrolytic solid made of zirconia-calcia or zirconia-yttria was disposed in a container. Liquid sodium containing oxygen flowed outside of the cylinder. Liquid sodium as a reducing material was present inside the cylinder and the container and the cylinder were electrically insulated. An electrode was inserted at the center of the cylinder and a baffle plate at the upper portion of the electrode to shield heat and rising sodium vapor was provided. The space above the container was filled with an inert gas. The oxygen in the liquid sodium to be purified transferred through the wall of the cylinder into the interior of the cylinder so as to oxydize the reducing sodium material. The supersaturated sodium oxide inside the cylinder was deposited.

  10. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

  11. Optimized evaporative cooling for sodium Bose-Einstein condensation against three-body loss

    International Nuclear Information System (INIS)

    Shobu, Takahiko; Yamaoka, Hironobu; Imai, Hiromitsu; Morinaga, Atsuo; Yamashita, Makoto

    2011-01-01

    We report on a highly efficient evaporative cooling optimized experimentally. We successfully created sodium Bose-Einstein condensates with 6.4x10 7 atoms starting from 6.6x10 9 thermal atoms trapped in a magnetic trap by employing a fast linear sweep of radio frequency at the final stage of evaporative cooling so as to overcome the serious three-body losses. The experimental results such as the cooling trajectory and the condensate growth quantitatively agree with the numerical simulations of evaporative cooling on the basis of the kinetic theory of a Bose gas carefully taking into account our specific experimental conditions. We further discuss theoretically a possibility of producing large condensates, more than 10 8 sodium atoms, by simply increasing the number of initial thermal trapped atoms and the corresponding optimization of evaporative cooling.

  12. TRU composition changes and their influence on FBR core characteristics in the LWR-to-FBR transition period

    International Nuclear Information System (INIS)

    Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

    2009-01-01

    In the conceptual core and fuel design studies in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan, much interest has been taken in the fuel nuclide compositions for a transition period from light water reactor to fast breeder reactor (FBR). In this paper, the range of transuranic (TRU) nuclide composition to be provided to FBR is evaluated with extended recycling scenario calculations. The influence of TRU composition changes on FBR core characteristics are also discussed with explanations of major contributing factors. (author)

  13. Reduction of HLW repository environmental impact by Na-Cooled FBR

    International Nuclear Information System (INIS)

    Ahn, Joonhong; Ikegami, Tetsuo

    2005-01-01

    The present study proposes the total toxicity index of radionuclides that have accumulated in the region exterior to the repository as the environmental impact measure. The measure is quantitatively evaluated by a radionuclide transport models that incorporate the effects of canister-array configuration and the initial mass loading in the waste canister. With the measure, it is demonstrated that the environmental impact of the repository can be effectively reduced by reduction of the initial mass loading resulting from FBR deployment and change in the canister-array configuration in the repository. The rate of increase in the environmental impact with the increase in the repository size can be reduced by reducing the initial mass loading of Np and its precursors. (author)

  14. FBR fuel cycle can be competitive

    International Nuclear Information System (INIS)

    Allardajs, R.; Kholl, R.; Pilling, R.

    1988-01-01

    The investigation results of comparative determination of averaged values for FBR and PWR reactor energy cost fuel components in England are described. Time periods till 2000 and 2020 yrs are considered. Dependence of the results obtained on changes of different initial factors is analysed. The FBR reactor fuel cost component is shown to be identically sensitive to the change is spent fuel cost. There is another condition for PWR reactors, where two factors are important: cost of source uranium ore concentrate and currency rate of exchange. The investigation carried out has shown that competitable fuel cost component of FBR reactors is reached even in first fast reactors, i.e. from the start of their large-scale construction. Thus, successive decrease of the capital component of energy cost is the key of success ful economical application of FBR reactors

  15. Sodium technology handbook

    International Nuclear Information System (INIS)

    2005-09-01

    This document was published as a textbook for the education and training of personnel working for operations and maintenances of sodium facilities including FBR plants and those engaged in R and D activities related to sodium technology. This handbook covers the following technical areas. Properties of sodium. Compatibilities of sodium with materials. Thermalhydraulics and structural integrity. Sodium systems and components. Sodium instrumentations. Sodium handling technology. Sodium related accident evaluation and countermeasures for FBRs. Operation, maintenance and repair technology of sodium facilities. Safety measures related to sodium. Laws, regulations and internal rules related to sodium. The plannings and discussions of the handbook were made in the Sodium Technology Education Committee organized in O-arai Engineering Center consisting of the representatives of the related departments including Tsuruga headquarters. Experts in various departments participated in writing individual technical subjects. (author)

  16. Experiment on a multilayer type air filter for the filtration of sodium aerosol

    International Nuclear Information System (INIS)

    Otake, N.; Nozaki, O.

    1987-01-01

    An emergency air filter system of FBR was developed by using a multilayer type filter to protect the function of HEPA filter from clogging due to loading of sodium aerosol. To examine the effect of loading of sodium aerosol on the filter system, sodium aerosol consisting of sodium oxides and the related compound was supplied to the filter system. Several parameters to determine the effectiveness of the multilayer type filter were surveyed. It was confirmed that the emergency air filter system of FBR consisting of the multilayer type filter, a medium filter, HEPA filter with standard size (610 mm x 610 mm) in series could hold 800 g-Na at 1.5 kPa without clogging

  17. Design and selection of materials for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.

    2011-01-01

    Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

  18. Study on design method for seismically isolated FBR plants

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Yabana, Shuichi; Ohtori, Yasuki; Ishida, Katsuhiko; Sawada, Yoshihiro; Shiojiri; Hiroo; Mazda, Taiji

    1998-01-01

    CRIEPI conducted 'Demonstration test on FBR seismic isolation system' from 1987 to 1996 under contract with Ministry of International Trade and Industry, Japan. In the demonstration test, base isolation technologies are prepared and demonstrated to apply to FBR and the design guidelines are proposed. In this report overall contents of the design guidelines entitled Design guidelines for seismically base isolated FBR plants' are included. The design guidelines, as a rule, are limited to apply to FBR plants where entire reactor building is isolated in the horizontal direction using laminated rubber bearings as isolators. The design guidelines and its concepts, however, will be useful for the development of similar guidelines for other isolation systems using different type of isolation methods and other nuclear facilities. The design guidelines consist of three parts and appendices. The first part is 'Policy for Safety Design of Base Isolated FBR Plants' specifying the principles and the requirements in the planning and the design for the safety of base isolated FBR plants. The second part is Policy for Seismic Design of Base Isolated FBR' describing the principles and the requirements in the seismic design and the evaluation of safety for base isolated FBR plants. The third part is 'Design Methods for Seismic Isolated FBR Plants' detailing the methods, procedures and parameters to be used in the design and the evaluation of safety fro base isolated FBR plants. In appendices examples of design procedures for base isolated reactor building and laminated rubber bearings as well as various test data on laminated rubber bearings, etc. are shown. (author)

  19. Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

    International Nuclear Information System (INIS)

    Althaus; Jansing; Kesseler; Kirchner; Menck

    1974-01-01

    Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

  20. Structure and short time degradation studies of sodium zirconium phosphate ceramics loaded with simulated fast breeder (FBR) waste

    Energy Technology Data Exchange (ETDEWEB)

    Ananthanarayanan, A., E-mail: arvinda@barc.gov.in [Process Development Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Ambashta, R.D., E-mail: aritu@barc.gov.in [Process Development Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sudarsan, V. [Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Ajithkumar, T. [Applied Catalysis Unit, National Chemical Laboratory, Pune 411008 (India); Sen, D.; Mazumder, S. [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Wattal, P.K. [Process Development Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2017-04-15

    Sodium zirconium phosphate (NZP) ceramics have been prepared using conventional sintering and hot isostatic pressing (HIP) routes. The structure of NZP ceramics, prepared using the HIP route, has been compared with conventionally sintered NZP using a combination of X-ray diffraction (XRD) and ({sup 31}P and {sup 23}Na) nuclear magnetic resonance (NMR) spectroscopy techniques. It is observed that NZP with no waste loading is aggressive toward the steel HIP-can during hot isostatic compaction and significant fraction of cations from the steel enter the ceramic material. Waste loaded NZP samples (10 wt% simulated FBR waste) show significantly low can-interaction and primary NZP phase is evident in this material. Upon exposure of can-interacted and waste loaded NZP to boiling water and steam, {sup 31}P NMR does not detect any major modifications in the network structure. However, the {sup 23}Na NMR spectra indicate migration of Na{sup +} ions from the surface and possible re-crystallization. This is corroborated by Small-Angle Neutron Scattering (SANS) data and Scanning Electron Microscopy (SEM) measurements carried out on these samples.

  1. Design study on simplification of secondary sodium cooling system for sodium cooled FBRs. Study result from JFY2000 to JFY2001

    International Nuclear Information System (INIS)

    Hori, Toru; Kawasaki, Nobuchika; Konomura, Mamoru

    2002-09-01

    For the 'Feasibility Studies on Commercialized Fast Reactor System' , various concepts with the simplified secondary sodium cooling system were designed, and the feasibility of technical issues was evaluated by focusing on improvement of economy and safety, especially elimination or mitigation of sodium-water direct interaction on heat transfer tube failure accident. In JFY 2000, 8 concepts with inert intermediate media were evaluated from standpoints of economy, safety, and structure integrity. And as promising candidates, the Pb-Bi pool type SG and the Pb-Bi tube type SG (concentric triple-walled tube) were selected, which had low cost compared with conventional IHX and SG system, and had potential of eliminating sodium-water direct interaction by separation of sodium and water tube zone. In JFY 2001, for the Pb-Bi tube type SG, important technical issues on 'Pb-Bi triple-walled tube specification suitable for safety demand', 'safety frame work corresponded to tube failure accident', and 'measures for Pb-Bi leakage into primary sodium loop' were studied, and the SG concept was constructed. In order to eliminate the design supposition of guillotine failure, available design measures for tube specification were tried to extract. But based on vibration characteristics of Pb-Bi triple-walled tube, the time required difference between outer and inner tube failure could not increase largely compared with known double-walled tube. The Pb-Bi tube type SG had potential of cost reduction (81% of cooling system, and 97% of plant), compared with conventional IHX and SG. But finally it was judged that design study on this type SG would not be executed after JFY 2002, due to impossibility of eliminating the design supposition of guillotine failure. (author)

  2. Sodium-cooled reactors, objectives, achieved technical state and development trends

    International Nuclear Information System (INIS)

    Wolff, U.

    1988-01-01

    The use of fossil fuels to cover the future world-wide energy demand alone would rapidly deplete these ressources, especially oil and gas. Today's knowledge suggests the enhanced exploitation of solar energy, nuclear fusion and the application of uranium in sodium-cooled breeder reactors as the alternative energies offering a great potential. The sodium-cooled reactor outdistances the other options in terms of development. Its technical feasibility and safe operation have been verified and its profitability appears to be possible when using today's technology. The verification of its profitability while maintaining a high safety level is the overriding task for the future. The paper discusses corresponding activities in the USA, the USSR, Japan and Western Europe. (orig.) [de

  3. LWR Spent Fuel Management for the Smooth Deployment of FBR

    International Nuclear Information System (INIS)

    Fukasawa, T.; Yamashita, J.; Hoshino, K.; Sasahira, A.; Inoue, T.; Minato, K.; Sato, S.

    2015-01-01

    Fast breeder reactors (FBR) and FBR fuel cycle are indispensable to prevent the global warming and to secure the long-term energy supply. Commercial FBR expects to be deployed from around 2050 until around 2110 in Japan by the replacement of light water reactors (LWR) after their 60 years life. The FBR deployment needs Pu (MOX) from the LWR-spent fuel (SF) reprocessing. As Japan can posses little excess Pu, its balance control is necessary between LWR-SF management (reprocessing) and FBR deployment. The fuel cycle systems were investigated for the smooth FBR deployment and the effectiveness of proposed flexible system was clarified in this work. (author)

  4. Dynamic thermal baffle on lower head of FBR sodium-sodium intermediate heat exchanger

    International Nuclear Information System (INIS)

    Charbonnel, A.; Foussat, C.

    1981-01-01

    The cover head of the heat exchanger is bathed on the one side by the primary sodium of the 'cold' header of the vessel and on the other side by the secondary sodium which feeds the heat exchange tube bank through the lower tubesheet. In the case of transient or permanent operating conditions at partial ratings, there are large temperature differences between the inner sodium (inlet temperature conditions of secondary sodium) and the outer sodium (mean temperature conditions in the primary sodium outlet port), hence the necessity of designing a thermal baffle which protects the head and its connection to the tubesheet. A 'static' thermal baffle consisting of a thick steel plate enclosing static sodium around the head proves inadequate during transient operating conditions. This is why a 'dynamic' thermal baffle is used whose design is based on the fact that the primary sodium in the lower part of the outlet port is always at a temperature close to that of the secondary sodium in the inlet header and the head. The primary sodium is taken from the bottom of the outlet port by a ring deflector and circulates in an annulus created by a double housing and the head. It flows out through openings in the lower part of the housing. (orig./GL)

  5. Physical properties of liquid sodium

    International Nuclear Information System (INIS)

    Alberdi Primicia, J.; Martinez Piquer, T.A.

    1977-01-01

    The molten sodium has been the more accepted coolant for the first generation of FBR, by this reason the knowledge of its technology is needed for the development of the next LMFBR. A series of necessary data for designing sodium liquid systems are given. Tables and graphics about the most important physical sodium properties between 1200-1400 degC are gathered. The results have been obtained from equations that relate the properties with temperature using a Fortran IV program. (author) [es

  6. Nitrogen gas extinguisher system as a countermeasure against a sodium fire at Monju

    International Nuclear Information System (INIS)

    Hasegawa, M.; Ikeda, M.; Kikuchi, H.

    2001-01-01

    Monju is a prototype sodium cooled FBR in Japan and occurred a sodium leakage incident in the secondary heat transport system on Dec. 8, 1995. The cause of the sodium leakage was a thermocouple well tube failure resulting from high cycle fatigue due to flow-induced vibration. The investigative research revealed that this type of flow-induced vibration was not a well-known Von Karman vortex shedding, but a symmetric vortex shedding. In the light of lessons from the sodium leakage incident, Monju will take several improvements in order to enhance the safety and reliability of the plant. A nitrogen gas extinguisher system will be installed at Monju as one of countermeasures against sodium fires. The basic design specifications of the system were determined by some experiments. Three kinds of experiment were conducted with the object of confirming; (1) an oxygen concentration to suppress the sodium fire, (2) a nitrogen gas mixing efficiency to decrease the oxygen concentration, and (3) a nitrogen gas feed rate to prevent air in-leak from the outside to keep the low oxygen atmosphere. This paper reports these tests which were performed to determine the design specification of the system. (authors)

  7. Nitrogen gas extinguisher system as a countermeasure against a sodium fire at Monju

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, M; Ikeda, M [MONJU Construction Office, Japan Nuclear Cycle Development Institute (Japan); Kikuchi, H [Kobe Shipyard, Mitsubishi Heavy Industries, Ltd, Kobe (Japan)

    2001-07-01

    Monju is a prototype sodium cooled FBR in Japan and occurred a sodium leakage incident in the secondary heat transport system on Dec. 8, 1995. The cause of the sodium leakage was a thermocouple well tube failure resulting from high cycle fatigue due to flow-induced vibration. The investigative research revealed that this type of flow-induced vibration was not a well-known Von Karman vortex shedding, but a symmetric vortex shedding. In the light of lessons from the sodium leakage incident, Monju will take several improvements in order to enhance the safety and reliability of the plant. A nitrogen gas extinguisher system will be installed at Monju as one of countermeasures against sodium fires. The basic design specifications of the system were determined by some experiments. Three kinds of experiment were conducted with the object of confirming; (1) an oxygen concentration to suppress the sodium fire, (2) a nitrogen gas mixing efficiency to decrease the oxygen concentration, and (3) a nitrogen gas feed rate to prevent air in-leak from the outside to keep the low oxygen atmosphere. This paper reports these tests which were performed to determine the design specification of the system. (authors)

  8. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  9. Thermal analysis experiment for elucidating sodium-water chemical reaction mechanism in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki

    2012-01-01

    For the purpose of elucidating the mechanism of the sodium-water surface reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature, and decomposition temperature of sodium hydride (NaH) have been identified from DTA curves. Based on the measured reaction temperature, rate constant of sodium monoxide (Na 2 O) generation was obtained. Thermal analysis results indicated that Na 2 O generation at the secondary overall reaction should be considered during the sodium-water reaction. (author)

  10. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-01-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  11. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Permana, Sidik; Suzuki, Mitsutoshi; Su' ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2012-06-06

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  12. Experimental Evaluation of Pool Fire Suppression Performance of Sodium Leak Collection Tray in Open Air

    International Nuclear Information System (INIS)

    Parida, F.C.; Rao, P.M.; Ramesh, S.S.; Malarvizhi, B.; Gopalakrishnan, V.; Rao, E.H.V.M.; Kasinathan, N.; Kannan, S.E.

    2006-01-01

    In the event of sodium leakage from heat transfer circuits of fast breeder reactors (FBR), liquid sodium catches fire in ambient air leading to production of flame, smoke and heat. One of the passive fire protection methods involves immediate collection of the leaking sodium to a sodium hold-up vessel (SHV) covered with a sloping cover tray (SCT) having a few drain pipes and one vent pipe (as in Fig. 1). As soon as the liquid sodium falls on the sloping cover tray, gravity guides the sodium through drain pipes into the bottom tray in which self-extinction occurs due to oxygen starvation. This sodium fire protection equipment called leak collection tray (LCT) works without the intervention of an operator and external power source. A large number of LCTs are strategically arranged under the sodium circulating pipe lines in the FBR plants to serve as passive suppression devices. In order to test the efficacy of the LCT, four tests were conducted. Two tests were with LCT having three drain pipes and rest with one. In each experiment, nearly 40 kg of hot liquid sodium at 550 deg. C was discharged on the LCT in the open air. Continuous on-line monitoring of temperature at strategic locations (∼ 28 points) were carried out. Colour video-graphy was employed for taking motion pictures of various time-dependent events like sodium dumping, appearance of flame and release of smoke through vent pipes. After self-extinction of sodium fire, the LCT was allowed to cool overnight in an argon atmosphere. Solid samples of sodium debris in the SCT and SHV were collected by manual core drilling machine. The samples were subjected to chemical analysis for determination of unburnt and burnt sodium. The results of the four tests revealed an interesting feature: LCT with three drain pipes showed far lower sodium collection efficiency and much higher sodium combustion than that with just one drain pipe. Thermal fluctuations in temperature sensor located near the tip of the drain pipe have

  13. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  14. Monte Carlo transport correction of sodium reactivity worth spatial distribution in perspective Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Raskach, K.F.; Blyskavka, V; Kislitsyna, T.S.

    2011-01-01

    In this paper we apply Monte Carlo for calculating spatial distribution of sodium reactivity worth in the perspective Russian sodium-cooled fast reactor BN-1200. A special Monte Carlo technique applicable for calculating perturbations and derivatives of the effective multiplication factor is used. The numerical results obtained show that Monte Carlo has a good perspective to deal with such problems and to be used as a reference solution for engineering codes based on the diffusion approximation. They also allow to conclude that in the sodium blanket and in the neighboring region of the core the diffusion code used likely overestimates sodium reactivity worth. This conclusion has to be verified in future work. (author)

  15. Summary: analysis of alternative FBR development strategies

    International Nuclear Information System (INIS)

    Burnham, J.B.

    1981-12-01

    This report summarizes the comparative evaluation of alternative strategies for the development of the commercial fast breeder reactor (FBR) in the United States. For planning purposes, a range of possible FBR development paths called strategies were selected for evaluation. These strategies, designed to be technically and economically feasible, were expressed in terms of the timing and nature of facilities/research and development programs required to reach full power operation of the first commercial FBR. Four of the seven strategies resulted in a large (1457 MWe) FBR as an end point, the other three in a 1000-MWe plant. Probability distributions were calculated for total strategy costs and time to completion. For the seven strategies analyzed, the costs (discounted 1980 dollars) ranged from $1.8 billion to $4.9 billion; the completion times ranged from 24 to 55 years

  16. FBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Tsugio.

    1986-01-01

    Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

  17. Development of system based code for integrity of FBR. Fundamental probabilistic approach, Part 1: Model calculation of creep-fatigue damage (Research report)

    International Nuclear Information System (INIS)

    Kawasaki, Nobuchika; Asayama, Tai

    2001-09-01

    Both reliability and safety have to be further improved for the successful commercialization of FBRs. At the same time, construction and operation costs need to be reduced to a same level of future LWRs. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of fabrication, installation, plant system design, safety design, operation and maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Failure probability is considered as a candidate index. Therefore we decided to make a model calculation using failure probability and judge its applicability. We first investigated other probabilistic standards like ASME Code Case N-578. A probabilistic approach in the structural integrity evaluation was created based on these results, and also an evaluation flow was proposed. According to this flow, a model calculation of creep-fatigue damage was performed. This trial calculation was for a vessel in a sodium-cooled FBR. As the result of this model calculation, a crack initiation probability and a crack penetration probability were found to be effective indices. Last we discussed merits of this System Based Code, which are presented in this report. Furthermore, this report presents future development tasks. (author)

  18. A reflux capsule steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Lantz, E.

    Pressurized water reactor plants at numerous sites have sustained significant leakage through their steam generators. The consequent shutdowns for repairs and replacements have damaged their economics. This experience suggests that if steam generators for liquid metal fast breeder reactors (LMFBR's) continue to be built as presently designed some of them will have similar problems. Because of their larger capital investment, the consequent damage to the economics of LMFBR's could be more serious. Reflux capsules provide a way to separate sodium from water and to reduce thermal stresses in steam generators for sodium cooled reactors. Their use would also eliminate the need for a primary heat exchanger and a secondary sodium loop pump. (author)

  19. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

  20. Transport of radioactive corrosion products in primary system of sodium-cooled fast breeder reactor 'MONJU'

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya

    2011-01-01

    Radioactive corrosion products (CP) are primary cause of personal radiation exposure during maintenance work at FBR plants with no breached fuel. The PSYCHE code has been developed based on the Solution-Precipitation model for analysis of CP transfer behavior. We predicted and analyzed the CP solution and precipitation behavior of MONJU to evaluate the applicability of the PSYCHE code to MONJU, using the parameters verified in the calculations for JOYO. From the calculation result pertaining to the MONJU system, distribution of 54 Mn deposited in the primary cooling system over 20 years of operation is predicted to be approximately 7 times larger than that of 60 Co. In particular, predictions show a notable tendency for 54 Mn precipitation to be distributed in the primary pump and cold-leg. The calculated distribution of 54 Mn and 60 Co in the primary cooling system of MONJU agreed with tendencies of measured distribution of JOYO. (author)

  1. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Ohashi, Yukio.

    1990-01-01

    The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

  2. Fatigue damage evaluation method for the longitudinal welded joint of a FBR main vessel in the vicinity of the sodium surface

    International Nuclear Information System (INIS)

    Tanigawa, Masayuki; Shimokoshi, Minoru; Negishi, Hitoshi; Nagata, Takashi

    1990-01-01

    Metallurgical discontinuities are dominant in the fatigue strength reductions at the welded joints of vessels whose surfaces could be finished. In the welded joints of SUS 304 with TYPE 308 weld metal fatigue strength reductions are caused by strain concentrations as the result of the softening of the weld metal. A combination model of two elastic fully plastic materials is applicable to the structures under thermal stresses where displacements are self-controlled. Metallurgical discontinuities are represented by the difference of the yield strength. The longitudinal welded joint of a large FBR main vessel in the vicinity of the sodium surface was analyzed using this model under various conditions related to the design. Strain concentrations at the welded joint could be evaluated using the elastic follow-up model. The maximum value of the elastic follow-up parameter was 3.0 if the yield stress ratio of the weld metal to the base metal was not less than 0.8. (author)

  3. The development and application of overheating failure model of FBR steam generator tubes. 2

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi

    2001-11-01

    The JNC technical report 'The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes' summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. 1. On the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. 2. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. 3. Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure. (author)

  4. Comparison of seismic isolation concepts for FBR

    International Nuclear Information System (INIS)

    Shiojiri, H.; Mazda, T.; Kasai, H.; Kanda, J.N.; Kubo, T.; Madokoro, M.; Shimomura, T.; Nojima, O.

    1989-01-01

    This paper seeks to verify the reliability and effectiveness of seismic isolation for FBR. Some results of the preliminary study of the program are described. Seismic isolation concepts and corresponding seismic isolation devices were selected. Three kinds of seismically-isolated FBR plant concepts were developed by applying promising seismic isolation concepts to the non-isolated FBR plant, and by developing plant component layout plans and building structural designs. Each plant was subjected to seismic response analysis and reduction in the amount of material of components and buildings were estimated for each seismic isolation concepts. Research and development items were evaluated

  5. Creep fatigue design of FBR components

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1997-01-01

    This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

  6. Development of the adjusted nuclear cross-section library based on JENDL-3.2 for large FBR

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki

    1999-04-01

    JNC (and PNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In 1991, the adjusted library based on JENDL-2, JFS-3-J2 (ADJ91R), was developed, and it has been used on the design research for FBR. As an evaluated nuclear library, however, JENDL-3.2 is recently used. Therefore, the authors developed an adjusted library based on JENDL-3.2 which is called JFS-3-J3.2(ADJ98). It is known that the adjusted library based on JENDL-2 overestimated the sodium void reactivity worth by 10-20%. It is expected that the adjusted library based on JENDL-3.2 solve the problem. The adjusted library JFS-3-J3.2(ADJ98) was produced with the same method as the adjusted library JFS-3-J2(ADJ91R) and used more integral parameters of JUPITER experiments than the adjusted library JFS-3-J2(ADJ91R). This report also describes the design accuracy estimation on a 600 MWe class FBR with the adjusted library JFS-3-J3.2(ADJ98). Its main nuclear design parameters (multiplication factor, burn-up reactivity loss, breeding ratio, etc.) except the sodium void reactivity worth which are calculated with the adjusted library JFS-3-J3.2(ADJ98) are almost the same as those predicted with JFS-3-J2(ADJ91R). As for the sodium void reactivity, the adjusted library JFS-3-J3.2(ADJ98) estimates about 4% smaller than the JFS-3-J2(ADJ91R) because of the change of the basic nuclear library from JENDL-2 to JENDL-3.2. (author)

  7. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    Ninokata, H.; Kamide, H.

    2011-01-01

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  8. Cost-benefit analysis on FBR cycle R and D for the world

    International Nuclear Information System (INIS)

    Kawasaki, Hirotsugu

    2006-01-01

    This analysis was estimated on the assumption that the nuclear power generation will be changed by FBR and both LWR and FBR indicate same nuclear power generation cost and the environmental load. The cost-benefit analysis results on FBR cycle R and D in the world showed that increase of power generation cost with increase of uranium fuel cost will be avoided and decrease of power generation cost by introducing FBR. The cost-benefit analysis results on FBR cycle R and D in Japan showed that about 9 billions yen will be obtained by the above two economic effects. Cost-benefit effects by introducing FBR, economic estimation method of cost-benefit effect, range and contents of cost-benefit effect on FBR R and D, preconditions of evaluation, and evaluation results are explained. (S.Y.)

  9. Development of FBR visual inspection technique in sodium

    International Nuclear Information System (INIS)

    Suzuki, T.; Nagai, S.; Shioyama, T.; Sato, M.; Karasawa, H.; Maruyama, F.; Ota, S.; Hida, T.; Kai, M.

    1995-01-01

    Because the reactor vessel of Fast Breeder Reactor is filled with opaque liquid sodium, it is expected to develop an acoustic visual inspection technique in sodium. The acoustic 3 dimensional image processing technique and the elemental parts of the visual inspection equipment in sodium have been developed at the first stage of the in-sodium visual inspection technique development. The cross correlation processing has been applied to improve the S/N ratio in the acoustic echo that are deteriorated by wetting in sodium and the low sensitivity that are also deteriorated by rather smaller diameter to integrate the high density multiple acoustic sensors. The improvement of S/N ratio has been realized by the cross correlation between acoustic echo data that is reflected from the objects and M-series continuous wave that is transmitted from the acoustic transducer. The high speed parallel processing circuits, in which DSPs (Digital Signal Processors) are included, have been developed to realize high speed processing by employing (the circuits connected to) each of sensors in parallel. Synthetic Aperture Focussing Technique (SAFT) has been applied to the acoustic 3-dimensional image processing. The amounts of ellipsoids must be drawn into the 3-dimensional memory to compose the 3-dimensional image by SAFT. Then, a high performance work-station has been employed to deal with enormous data to compose the acoustic 3 dimensional image. Motors and cables, which can be operated under the condition of high-temperature and high-radiation environment, have been developed as the parts of the manipulator which will be used for visual inspection equipment in sodium. A prototype drive mechanism consists of the manipulator with three joints and a scanner with an arrayed acoustic sensors which a sweeps in fan-shape mechanically. The manipulatory type prototype drive mechanism and the signal processing device have been developed and tested, and the acoustic 3-dimensional image of pyramid

  10. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  11. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  12. Challenges in licensing a sodium-cooled advanced recycling reactor

    International Nuclear Information System (INIS)

    Levin, Alan E.

    2008-01-01

    As part of the Global Nuclear Energy Partnership (GNEP), the U.S. Department of Energy (DOE) has focused on the use of sodium-cooled fast reactors (SFRs) for the destruction of minor actinides derived from used reactor fuel. This approach engenders an array of challenges with respect to the licensing of the reactor: the U.S. Nuclear Regulatory Commission (NRC) has never completed the review of an application for an operating license for a sodium-cooled reactor. Moreover, the current U.S. regulatory structure has been developed to deal almost exclusively with light-water reactor (LWR) designs. Consequently, the NRC must either (1) develop a new regulatory process for SFRs, or (2) reinterpret the existing regulations to apply them, as appropriate, to SFR designs. During the 1980s and 1990s, the NRC conducted preliminary safety assessments of the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Innovative Small Module (PRISM) designs, and in that context, began to consider how to apply LWR-based regulations to SFR designs. This paper builds on that work to consider the challenges, from the reactor designer's point of view, associated with licensing an SFR today, considering (1) the evolution of SFR designs, (2) the particular requirements of reactor designs to meet GNEP objectives, and (3) the evolution of NRC regulations since the conclusion of the SAFR and PRISM reviews. (author)

  13. Development Status on Innovative Sodium-Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Yanagisawa, Tsutomu; Sato, Kazujiro

    2006-01-01

    The first step in Japan's nuclear fuel cycle policy is to introduce MOX recycle in light water reactors (LWRs) and the final step is to establish multiple TRU recycle in fast reactors (FRs), with the goal of realizing a stable supply, effective use of nuclear fuel resources, and the environmentally friendly production of energy. Therefore, a feasibility study on commercialized FR cycle systems has been launched since July 1999 by a Japanese joint project team of Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC: the representative of the electric utilities) in cooperation with Central Research Institute of Electric Power Industry (CRIEPI) and vendors. In the period from July 1999 to March 2001, the feasibility study phase-I was conducted to screen out representative FR cycle concepts. In the feasibility study phase-II (April 2001 - March 2006), investigations in to the representative FR concepts were carried out to clarify the most promising concept for commercial deployment. This paper describes an innovative sodium-cooled FR, which is named as the JAEA Sodium-cooled FR (JSFR), as the most promising FR concept that meets the Generation-IV performance target. The JSFR employs several advanced technologies, such as an oxide dispersion strengthened (ODS) cladding for higher burn-up, a short-piping configuration with less elbows by adopting high chromium steel, a large scale integrated intermediate heat exchanger with a primary circulation pump, etc. Based on the design, construction and operation experiences of JOYO and MONJU, there are extensive technology bases for sodium-cooled FRs. Nevertheless, several innovative technologies implemented into the JSFR have to be developed in order to realize higher economic competitiveness by reducing construction costs and improving plant availability

  14. Roof slab cooling device in a FBR type reactor

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1987-01-01

    Purpose: To obtain a roof slab cooling device capable of retaining cooling performance even in a case of electric power supply stop or failure and effective from economical point of view. Constitution: Atmospheric air is introduced into the cooling chamber of a proof slab and spontaneously passed to a exit pipeway connected to a stack thereby cooling the roof slab. Specifically, atmospheric air entered from the inlet pipeway is introduced to the cooling chamber and absorbs heat generate from the inside of the reactor container. Warmed air is sucked from the exit pipeway and then released into the atmosphere passing through the stack. The air cools the roof slab during circulation due to spontaneous passage and keeps the slab at a low temperature. Since the air is passed spontaneously, no power such as for a blower is required at all and, if the electric power supply should be lost, the cooling power can be maintained as it is to provide a high reliability. Further, since no electric power is required for the blowing power, it has high economical merit. (Horiuchi, T.)

  15. Survey of evaluation methods for thermal striping in FBR structures

    International Nuclear Information System (INIS)

    Miura, Naoki; Nitta, Akito; Take, Kohji

    1988-01-01

    In the upper core structures or the sodium mixing tee of Fast Breeder Reactors, sodium mixing streams which are at different temperatures produce rapid temperature fluctuations, namely 'thermal striping', upon component surfaces, and it is apprehended that the high-cycle thermal fatigue causes the crack initiation and propagation. The thermal striping is one of the factors which is considered in FBR component design, however, the standard evaluation method has not built up yet because of the intricacy of that mechanism, the difficulty of an actual proof, the lack of data, and so on. In this report, it is intended to survey of the datails and the present situation of the evaluation method of crack initiation and propagation due to thermal striping, and study the appropriate method which will be made use of the rationalization of design. So it is ascertained that the method which use a quantitative prediction of crack propagation is optimum to evaluate the thermal striping phenomenon. (author)

  16. Design considerations for economically competitive sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

    2009-01-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

  17. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  18. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    International Nuclear Information System (INIS)

    Song, Wei; Wu, Yuanyu; Hu, Wenjun; Zuo, Jiaxu

    2015-01-01

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  19. 4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

    International Nuclear Information System (INIS)

    2012-12-01

    The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

  20. An evaluation procedure of sodium environmental effects on FBR grade SUS316 (316FR) and Modified 9Cr-1Mo steel. On the basis of the studies up to the fiscal year of 1998

    International Nuclear Information System (INIS)

    1999-01-01

    Evaluation of sodium environmental effects on structural materials of fast breeder reactors (FBR's) is one of the key issues for the integrity of the plants. The Elevated Temperature Structural Design Guide for Monju (ETSDG) incorporated an evaluation procedure of sodium environmental effects in the Appendix MA.2, for the conventional steels, such as SUS304, SUS316, SUS321 and 2 1/4Cr-1Mo. Following the establishment of the ETSDG, a new material with superior elevated temperature properties, FBR grade SUS316 (316FR), has been developed, and studies on Mod.9Cr-1Mo steel (Mod.9Cr-1Mo steel) has been performed, for the application to demonstration reactors and successive large-scale reactors. These materials were shown to have, at least equal, or better compatibility with sodium compared with the conventional steels. Moreover, studies have been continued with the conventional steels, particularly with SUS304, for the further validation of the procedure in the ETSDG, especially in terms of long-term properties. Those studies provide basis for the study on 316FR. This report proposed an evaluation procedure of sodium environmental effects on 316FR and Mod.9Cr-1Mo steel, which is to be incorporated into the structural design guide for demonstration fast breeder reactors. The procedure is summarized as follows: (1) Corrosion allowance of 316FR and Mod.9Cr-1Mo can be evaluated by the equation determined in the ETSTG. (2) Strength reduction factors on design allowable values are not necessary for either steel. Strength reduction due to the transfer of carbon and nitrogen, etc does not occur with 316FR, which was the same as SUS304. Mod.9Cr-1Mo steel does not show strength reduction, contrary to 2 l/4Cr-1Mo, similar ferritic steel. (3) Corrosion allowance can be determined separately for thin-walled components. The procedure allows design without correction factors for Mod.9Cr-1Mo steel, which was not possible for 2 1/4Cr-1Mo steel in the ETSDG. (author)

  1. Success tree analysis on the technologies development for FBR commercialization

    International Nuclear Information System (INIS)

    An, Shigehiro; Taniyama, Hiroshi; Nagai, Hiroshi.

    1991-01-01

    In order to obtain a secure energy supply in future, it is important to establish a system for plutonium utilization via the FBR which is superior to the uranium utilization system with respect to both safety and good economics. In spite of this obvious need, the commercialization of the FBR is facing delays. Although several factors, for example, improvement of LWR technologies, stable supply of low cost uranium, opposition to nuclear power, etc. are contributors, the primary reason for the delay is the unfavorable economics of the FBR itself. In this paper the key technologies leading to reduced FBR costs are identified and their development strategies are discussed. (author)

  2. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  3. Flexible fuel cycle system for the transition from LWR to FBR

    International Nuclear Information System (INIS)

    Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Sasahira, Akira; Inoue, Tadashi; Minato, Kazuo; Sato, Seichi

    2009-01-01

    Japan will deploy commercial fast breeder reactor (FBR) from around 2050 under the suitable conditions for the replacement of light water reactor (LWR) with FBR. The transition scenario from LWR to FBR is investigated in detail and the flexible fuel cycle initiative (FFCI) system has been proposed as a optimum transition system. The FFCI removes ∼95% uranium from LWR spent fuel (SF) in LWR reprocessing and residual material named Recycle Material (RM), which is ∼1/10 volume of original SF and contains ∼50% U, ∼10% Pu and ∼40% other nuclides, is treated in FBR reprocessing to recover Pu and U. If the FBR deployment speed becomes lower, the RM will be stored until the higher speed again. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR SF during the transition period. The economy is better for FFCI due to the smaller LWR reprocessing facility (no Pu/U recovery and fabrication). The FFCI can supply high Pu concentration RM, which has high proliferation resistance and flexibly respond to FBR introduction rate changes. Volume minimization of LWR SF is possible for FFCI by its conversion to RM. Several features of FFCI were quantitatively evaluated such as Pu mass balance, reprocessing capacities, LWR SF amounts, RM amounts, and proliferation resistance to compare the effectiveness of the FFCI system with other systems. The calculated Pu balance revealed that the FFCI could supply enough but no excess Pu to FBR. These evaluations demonstrated the applicability of FFCI system to the transition period from LWR to FBR cycles. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  5. Development of tube rupture evaluation code for FBR steam generator (II). Modification of heat transfer model in sodium side

    International Nuclear Information System (INIS)

    Hamada, H.; Kurihara, A.

    2003-05-01

    The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1) The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in comparison with the previous method. (2) The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3) The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4) In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5) In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model. (author)

  6. Conceptual design for accelerator-driven sodium-cooled sub-critical transmutation reactors using scale laws

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates the reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scaled and verified through the methodology in this paper, which is referred to Advanced Liquid Metal Reactor (ALMR). A Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the integrated code system. Integrated Code System (ICS) consists of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related regions are analyzed once-through Results of conceptual design are attached in this paper. 5 refs., 4 figs., 1 tab. (Author)

  7. Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

  8. Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

    International Nuclear Information System (INIS)

    Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum

    2005-01-01

    KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types

  9. Study on the FBR cycle introduction scenario. 4. Evaluation of the FBR cycle introduction scenario from the viewpoints of the fuel cycle requirements

    International Nuclear Information System (INIS)

    Ono, Kiyoshi; Shiotani, Hiroki; Hirao, Kazunori

    2003-07-01

    This report is intended to explain the outline of the scenario studies on FBR (Fast Breeder Reactor) cycle introduction. Recently, people value the reduction of environmental impact in addition to the recycle of energy resources and the energy security in these scenario studies. This report summarizes the analysis about the necessity of plutonium recycling in LWR (Light water Reactor) from short-term view and about the necessity of FBR cycle introduction from a long-term view in Japan, by comparing 'FBR scenario' with 'LWR once-through scenario' and 'Pu recycle in LWR scenario', from the viewpoints of cumulative uranium demand, spent fuel storage, radioactive waste arising, etc. It becomes clear that the plutonium recycling in LWR has a good effect on the reduction of spent fuel storage and the cumulative natural uranium demand before FBR cycle introduction, from short-term view (20-30 years). On the other hand, this analysis also shows that there is much effect of FBR deployment not only on saving amount of uranium use and energy security but also on reduction of high-level radioactive waste (spent fuels and vitrified waste) and minor actinide arising, from long-term view (100-200 years). (author)

  10. Computational methodology of sodium-water reaction phenomenon in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Takata, Takashi; Yamaguchi, Akira; Uchibori, Akihiro; Ohshima, Hiroyuki

    2009-01-01

    A new computational methodology of sodium-water reaction (SWR), which occurs in a steam generator of a liquid-sodium-cooled fast reactor when a heat transfer tube in the steam generator fails, has been developed considering multidimensional and multiphysics thermal hydraulics. Two kinds of reaction models are proposed in accordance with a phase of sodium as a reactant. One is the surface reaction model in which water vapor reacts directly with liquid sodium at the interface between the liquid sodium and the water vapor. The reaction heat will lead to a vigorous evaporation of liquid sodium, resulting in a reaction of gas-phase sodium. This is designated as the gas-phase reaction model. These two models are coupled with a multidimensional, multicomponent gas, and multiphase thermal hydraulics simulation method with compressibility (named the 'SERAPHIM' code). Using the present methodology, a numerical investigation of the SWR under a pin-bundle configuration (a benchmark analysis of the SWAT-1R experiment) has been carried out. As a result, the maximum gas temperature of approximately 1,300degC is predicted stably, which lies within the range of previous experimental observations. It is also demonstrated that the maximum temperature of the mass weighted average in the analysis agrees reasonably well with the experimental result measured by thermocouples. The present methodology will be promising to establish a theoretical and mechanical modeling of secondary failure propagation of heat transfer tubes due to such as an overheating rupture and a wastage. (author)

  11. Study of an electromagnetic pump in a sodium cooled reactor. Design study of secondary sodium main pumps (Joint research)

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Kisohara, Naoyuki; Hishida, Masahiko; Fujii, Tadashi; Konomura, Mamoru; Ara, Kuniaki; Hori, Toru; Uchida, Akihito; Nishiguchi, Youhei; Nibe, Nobuaki

    2006-07-01

    In the feasibility study on commercialized fast breeder cycle system, a medium scale sodium cooled reactor with 750 MW electricity has been designed. In this study, EMPs are applied to the secondary sodium main pump. The EMPs type is selected to be an annular linear induction pump (ALIP) type with double stators which is used in the 160 m 3 /min EMP demonstration test. The inner structure and electromagnetic features are decided reviewing the 160 m 3 /min EMP. Two dimensional electromagnetic fluid analyses by EAGLE code show that Rms (magnetic Reynolds number times slip) is evaluated to be 1.08 which is less than the stability limit 1.4 confirmed by the 160 m 3 /min EMP test, and the instability of the pump head is evaluated to be 3% of the normal operating pump head. Since the EMP stators are cooled by contacting coolant sodium duct, reliability of the inner structures are confirmed by temperature distribution and stator-duct contact pressure analyses. Besides, a power supply system, maintenance and repair feature and R and D plan of EMP are reported. (author)

  12. Startup of the FFTF sodium cooled reactor

    International Nuclear Information System (INIS)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed

  13. Developing works to detect fatigue cracks (small sodium leak detector and acoustic emission

    International Nuclear Information System (INIS)

    Kikuchi, M.; Sakakibara, Y.; Nagata, T.

    1980-01-01

    Continuous monitoring of fatigue cracks was performed (using both sodium leak detector and AE measuring system) through the creep-fatigue test of 304 stainless steel long elbow as part of the test series to establish the structural reliability of the Prototype FBR primary heat transport piping system. The sodium leak detector was a system composed mainly of SID (Sodium Ionization Detector) and DPD (Deferential Pressure Detector), that was developed by HITACHI Ltd. under a contract with PNC. The AE system was Synthetic AE Measuring and Analyzing system that was developed at FBR Safety Section to measure and analyze AE at various piping component tests. The test was continued until a sodium leakage was detected by the contact-type sodium leak detector attached to the test assembly, after about 4 weeks operation under cyclic loading at 600 deg. C. The following conclusions were obtained: (1) The sodium leak detector, both SID and DPD, indicated sodium leakage clearly, some hours before the contact-type detector did, even under an environment of air that contains ordinary humidity (Leaked sodium was estimated to be less than 15 grams after completion of the test); (2) The AE method indicated location and seriousness of the fatigue cracks, apparently before the crack penetration occurred. (author)

  14. Part 6. Internationalization and collocation of FBR fuel cycle facilities

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Abramson, P.B.; LeSage, L.G.

    1980-01-01

    This report examines some of the non-proliferation, technical, and institutional aspects of internationalization and/or collocation of major facilities of the Fast Breeder Reactor (FBR) fuel cycle. The national incentives and disincentives for establishment of FBR Fuel Cycle Centers are enumerated. The technical, legal, and administrative considerations in determining the feasibility of FBR Fuel Cycle Centers are addressed by making comparisons with Light Water Reactor (LWR) centers which have been studied in detail by the IAEA and UNSRC

  15. Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Saito, Masaki; Novitrian,; Waris, Abdul; Suud, Zaki

    2013-01-01

    Highlights: • The paper analyzes the plutonium production of recycling nuclear fuel option. • To evaluate material attractiveness based on intrinsic feature of material barrier. • Evaluation based on isotopic plutonium composition of spent fuel LWR and FBR. • Even mass number of plutonium gives a significant contribution to material barrier, in particular Pu-238 and Pu-240. • Doping MA in FBR blanket is effective to increase material barrier from weapon grade plutonium to more than MOX fuel grade. - Abstract: Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no

  16. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  17. Accurate reactivity void coefficient calculation for the fast spectrum reactor FBR-IME

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Fabiano P.C.; Vellozo, Sergio de O.; Velozo, Marta J., E-mail: fabianopetruceli@outlook.com, E-mail: vellozo@cbpf.br, E-mail: martajann@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Militar

    2017-07-01

    This paper aims to present an accurate calculation of the void reactivity coefficient for the FBR-IME, a fast spectrum reactor in development at the Engineering Military Institute (IME). The main design peculiarity lies in using mixed oxide [MOX - PuO{sub 2} + U(natural uranium)O{sub 2}] as fuel core. For this task, SCALE system was used to calculate the reactivity for several voids distributions generated by bubbles in the sodium beyond its boiling point. The results show that although the void reactivity coefficient is positive and location dependent, they are offset by other feedback effects, resulting in a negative overall coefficient. (author)

  18. Characterization of alternative FBR development strategies

    International Nuclear Information System (INIS)

    Boegel, A.J.; Clausen, M.J.

    1981-08-01

    Near-term decisions regarding the nature and place of the FBR development program must be made. This study is part of a larger program designed to provide the Department of Energy (DOE) with imformation that can be used to make strategic programmatic decisions. The focus of this report is the description of alternative approaches for developing the FBR and the quantification of the duration and cost of each alternative. The time frames of the alternative approaches are investigated in companion reports (White 1981 and Fraley 1981). The results of these analyses will be described in a summary report

  19. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2015-01-01

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

  20. Development of FBR cycle data base system (II)

    International Nuclear Information System (INIS)

    Kubota, Sadae; Ohtaki, Akira; Hirao, Kazuhiro

    2003-05-01

    In the 'Feasibility Study on Commercialized FBR Cycle Systems (F/S)', scenario evaluations, cost-benefit evaluations and system characteristic evaluations to show the significance of the FBR cycle system introduction concretely are performed while design studies for FBR plants, reprocessing systems and fabrication systems are conducted. In these evaluations, future society of various conditions and situation is assumed, and investigation and analysis about needs and social effects of FBR cycle are carried out. In this study, promising FBR cycle concepts are suggested by taking information such as domestic and foreign policies and bills, an economic prediction, a supply and demand prediction of resources, a project of technology development into consideration in addition to system design information. The development of the FBR Cycle Database which this report introduced started in 1999 fiscal year to enable managed unitarity and searched reference information to use for the above scenario evaluations, cost-benefit evaluations and system characteristic evaluations. In 2000 fiscal year, its prototype was made and used tentatively, and we extracted the problems in operation and functions from that, and, in 2001 fiscal year, the entry system and the search system using the Web page were made in order to solve problems of the prototype, and started use in our group. Moreover, in 2002 fiscal year, we expanded and improved the search system and promoted the efficiency of management work, and use in JNC through intranet of the database was started. In addition, as a result of having made the entry of about 350 data in 2002 fiscal year, the collected number of the database reaches about 7,250 by the end of March, 2003. We are to continue the entry of related information of various evaluations in F/S phase 2 from now on. In addition, we are to examine improvement of convenience of the search system and cooperation with the economy database. (author)

  1. Development of FBR piping bellows joint

    International Nuclear Information System (INIS)

    Tsukimori, Kazuyuki; Iwata, Koji

    1991-01-01

    Reduction of construction cost is one of the most important problems to realize a FBR (Fast Breeder Reactor) Plant. Significant reduction of the construction cost of a reactor building, related equipments and facilities can be expected by shortening the length of its long cooling pipes. Since the bellows has a great capacity for absorbing thermal expansion displacement, application of bellows expansion joints is considered as the most influential measure for reduction of the piping length. To confirm technological possibilities of application and practical use of bellows joints in the main piping systems, extensive R and D's, development of various methods for evaluating the strength of bellows, establishment of inspection and maintenance techniques, studies on safety logic, etc., were carried out by PNC from 1983 to 1988. Through these studies, technological possibilities of bellows joints were confirmed and the results were summarized in the 'Structural Design Guide for Class 1 Piping Bellows Expansion Joints of Fast Breeder Reactor for Elevated Temperature Service' and the 'Inspection and Maintenance Standards of Piping bellows expansion Joints'. (author)

  2. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors

    International Nuclear Information System (INIS)

    Vega R, A. K.; Espinosa P, G.; Gomez T, A. M.

    2016-09-01

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  3. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

    2013-01-01

    • Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

  4. Operational reliability testing of FBR fuel in EBR-II

    International Nuclear Information System (INIS)

    Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

    1991-01-01

    The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

  5. Some new fatigue tests in high temperature water and liquid sodium environment

    International Nuclear Information System (INIS)

    Hattori, Takahiro; Yamauchi, Takayoshi; Kanasaki, Hiroshi; Kondo, Yoshiyuki; Endo, Tadayoshi.

    1987-01-01

    To evaluate the fatigue strength of structural materials for PWR or FBR plants, fatigue test data must be obtained in an environment of simulated primary and secondary water for PWR or of high temperature liquid sodium for FBR. Generally, such tests make it necessary to prepare expensive facilities, so when large amount of fatigue data are required, it is necessary to rationalize and simplify the fatigue tests while maintaining high accuracy. At the Takasago Research Development Center, efforts to rationalize facilities and maintain accuracy in fatigue tests have been made by developing new test methods and improving conventional techniques. This paper introduces a new method of low cycle fatigue test in high temperature water, techniques for automatic measurement of crack initiation and propagation in high temperature water environment and a multiple type fatigue testing machine for high temperature liquid sodium. (author)

  6. Development of FBR cycle data base system

    International Nuclear Information System (INIS)

    Kubota, Sadae; Ohtaki, Akira; Hirao, Kazuhiro

    2002-06-01

    In the 'Feasibility Study on Commercialized Fast Reactor Cycle System (F/S)'. scenario evaluations, cost-benefit evaluations and system characteristic evaluations to show significance of the Fast Breeder Reactor (FBR) cycle system introduction concretely are performed in parallel with a design study for FBR plants, reprocessing systems and fabrication systems. In these evaluations, informations such as economic prospects, prospects for supply and demand of resources and a progress of engineering development are used in addition to design information. This report explains a FBR Cycle Database in order to carry out management and search of various design information and the relating information. The prototype system of the database was completed in the 2000 fiscal year, and the problem of the user number restriction of the prototype system has been improved by Web-ization in the 2001 fiscal year. About 7,000 data are stored in this data base (as of the end of March, 2002). The expansion of user etc., and the continuation of input work of various evaluation information will be carried out, in the phase 2 of F/S. (author)

  7. Passive safety optimization in liquid-sodium cooled reactors

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hahn, D.; Chang, W.-P.; Kwon, Y.-M.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2004-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4)

  8. Design of sodium cooled reactor systems and components for maintainability

    International Nuclear Information System (INIS)

    Carr, R.W.; Charnock, H.O.; McBride, J.P.

    1978-09-01

    Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

  9. Final report-passive safety optimization in liquid sodium-cooled reactors

    International Nuclear Information System (INIS)

    Cahalana, J. E.; Hahn, D.

    2007-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  10. Studies of decay heat removal by natural convection using the SONACO sodium-cooled 37-pin bundle

    International Nuclear Information System (INIS)

    Wydler, P.; Dury, T.V.; Hudina, M.; Weissenfluh, T. von; Sigg, B.; Dutton, P.

    1986-01-01

    Natural convection measurements in an electrically heated sodium-cooled rod bundle are being performed with the aim of contributing to a better understanding of natural convection effects in subassemblies with stagnant sodium and providing data for code validation. Measurements include temperature distributions in the bundle for different cooling configurations which simulate heat transfer to the intersubassembly gap and neighbouring subassemblies and possible thermosyphonic interaction between a subassembly and the reactor plenum above. Conditions for which stable natural convection patterns exist are identified, and results are compared with predictions of different computer codes of the porous-medium type. (author)

  11. Sodium flow measurement in large pipelines of sodium cooled fast breeder reactors with bypass type flow meters

    International Nuclear Information System (INIS)

    Rajan, K.K.; Jayakumar, T.; Aggarwal, P.K.; Vinod, V.

    2016-01-01

    Highlights: • Bypass type permanent magnet flow meters are more suitable for sodium flow measurement. • A higher sodium velocity through the PMFM sensor will increase its sensitivity and resolution. • By modifying the geometry of bypass line, higher sodium velocity through sensor is achieved. • With optimized geometry the sensitivity of bypass flow meter system was increased by 70%. - Abstract: Liquid sodium flow through the pipelines of sodium cooled fast breeder reactor circuits are measured using electromagnetic flow meters. Bypass type flow meter with a permanent magnet flow meter as sensor in the bypass line is selected for the flow measurement in the 800 NB main secondary pipe line of 500 MWe Prototype Fast Breeder Reactor (PFBR), which is at the advanced stage of construction at Kalpakkam. For increasing the sensitivity of bypass flow meters in future SFRs, alternative bypass geometry was considered. The performance enhancement of the proposed geometry was evaluated by experimental and numerical methods using scaled down models. From the studies it is observed that the new configuration increases the sensitivity of bypass flow meter system by around 70%. Using experimentally validated numerical tools the volumetric flow ratio for the bypass configurations is established for the operating range of Reynolds numbers.

  12. Simulation of radioactive corrosion product in primary cooling system of Japanese sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    2012-01-01

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54 Mn and 60 Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54 Mn was estimated to constitute approximately 20% and 60 Co approximately 40% in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO. (author)

  13. Materials and manufacturing for sodium cooled breeder and fusion power reactor

    International Nuclear Information System (INIS)

    Baldev Raj

    2013-01-01

    The paper narrates definitions of challenges relating to materials and manufacturing for sodium cooled fast reactors thermonuclear fusion reactors. Science and technology developed indigenously but in the context of bench marks in the world is described through examples. Solutions to challenges requires synergy among theoretical physicists, computational chemists, material scientists, metallurgists and engineers with their domains of expertise along with foresight effective management

  14. Heavy liquid metal cooled FBR. Results 2003

    International Nuclear Information System (INIS)

    Hayahune, Hiroki; Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu

    2004-08-01

    Concepts of the reactor, SG and main coolant pump have been studied considering maintainability and aseismic capability, which is a medium size pool type lead-bismuth cooled reactor. The results are following. (1) Reconsideration of reactor design concepts concerning maintainability: In pursuit of good reactor maintainability, the structural concepts of SG, UIS and core support structures have been changed to be drawn up above the upper area of the reactor system. After a few decade of interval, lead-bismuth inventory in the reactor vessel shall be fully drained for easy ISI operation of in-vessel main components such as core support structures. From the viewpoint of the reactor aseismic capability, the axial length of reactor vessel was reduced and the reactor vessel support location was changed from the top handing to the circumference of the vessel. (2) SG concept selection in conjunction with a compact reactor vessel: The concept of SG consisting of a once through type with helical coil tube is selected. 6 units of a small scale SG are arranged on a reactor roof deck along the peripheral direction, in addition to 3 units of a centrifugal mechanical pump. (3) Aseismic structural integrity of the reactor components: Aseismic structural integrity of the reactor vessel, core support structures, UIS, FHM, SG and the main pumps has been vigorously examined respectively. These components besides FHM could keep the aseismic structural integrity for strong S2 earthquake under the design condition. FHM could also keep the integrity for S1 earthquake. (4) Safety evaluation: Thermal transients following loss of flow type accident due to plant total blackout and typical manual reactor trip incident, have been evaluated to assure the pant safety design, by analyzing thermal hydraulic behavior of transients concerning core flow rate and temperatures of the plant cooling system. Loss of flow accident due to plant total blackout: The reactor coolant pumps shall be tripped and

  15. Application of seismic isolation technology to demonstration FBR

    International Nuclear Information System (INIS)

    Kato, Muneaki

    1994-01-01

    The Japanese demonstration FBR is loop type, the intermediate heat exchanger is installed between the reactor and the steam generator, and up to the intermediate heat exchanger is in the containment vessel, which is designed as a reinforced concrete vessel. In FBRs, the optimization in aseismatic design and high temperature structural design is important. The reactor building is buried in rock bed up to its center of gravity to minimize the amplifying earthquake response. If the seismic isolation structure for a reactor is realized, cost reduction can be expected by the rationalization of machinery and equipment and the standardization of buildings and facilities. The research on FBR seismic isolation design has been carried out by Central Research Institute of Electric Power Industry and Japan Atomic Power Co. The concept of FBR seismic isolation design, the basic condition for the design evaluation, the research on safety allowance and the conceptual design analysis are reported. (K.I.)

  16. Selection of steam generator materials for sodium cooled fast breeders

    International Nuclear Information System (INIS)

    Berge, P.

    1977-01-01

    The sodium water heat exchangers are now considered as the stumbling block in the development of liquid metal cooled fast breeders, due to the risk of sodium-water reactions. The selection of the materials for these tube-bundles has been very broad, for the different existing, or in-project, reactors in the world: low alloy 2 1/4 Cr - 1 Mo steels (unstabilized or stabilized); 9 Cr - 1 Mo ferritic steel; 18 Cr - 10 Ni austenitic stainless steels; alloy 800. On can also add other ferritic steels, as 9 Cr - 2 Mo stabilized, which are studied for this application. In the framework of the E.D.F.-C.E.A. working group a major effort was undertaken to study the characteristics of these various materials with respect to the main criteria governing construction of the tube bundles and their performance in service: mechanical characteristics at high temperature; fabrication and welding; behavior with respect to mass transfer in sodium; carburization and decarburization; corrosion resistance. The main lines and results of this program are described [fr

  17. Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2015-02-15

    Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  18. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  19. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, H.; Eoh, J.; Cha, J.; Kim, S.

    2011-01-01

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  20. Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

    International Nuclear Information System (INIS)

    You, WuSeung; Hong, Ser Gi

    2014-01-01

    The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

  1. New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2017-07-15

    The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

  2. Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

    International Nuclear Information System (INIS)

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2009-01-01

    Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

  3. Temperature monitoring and leak detection in sodium circuits of FBR using Raman distributed fiber optic sensor

    International Nuclear Information System (INIS)

    Kasinathan, M.; Murali, N.; Sosamma, S.; Babu Rao, C.; Kumar, Anish; Purnachandra Rao, B.; Jayakumar, T.

    2013-01-01

    This paper discusses the fiber optic temperature sensor based leak detection in the coolant circuits of fast breeder reactor. These sensors measure the temperature based on spontaneous Raman scattering principle and is not influenced by the electromagnetic interference. Various experiments were conducted to evaluate the performance of the fiber optic sensor based leak detection using Raman distributed Temperature Sensor (RDTS). This paper also deals with the details of fiber optic sensor type leak detector layout for the coolant circuit of FBR, performance requirement of leak detection system, description of the test facility, experimental procedure and test results of various experiments conducted. (author)

  4. Technological study report on synthetic evaluation for FBR cycle. The report of the feasibility studies on commercialized FBR cycle system. Phase 1

    International Nuclear Information System (INIS)

    Shinoda, Yoshihiko; Ohtaki, Akira; Kofuji, Hirohide; Ono, Kiyoshi; Hirao, Kazunori

    2001-03-01

    This report is intended to explain the outline of the characteristic evaluation work on various FR cycle system concepts, following the design work, in the 1st phase of the JNC's 'Feasibility Study on Commercialized Fast Reactor Cycle System (the F/S)' (from 1999 to March 2001). The purpose of this characteristic evaluation is to reveal the performance of candidate FR cycle systems. For this synthetic estimation, six viewpoints, such as Economics, Effective utilization of uranium resource, Reduction of environmental impact, Safety, Proliferation resistance, and Technological feasibility, are selected. In addition, aiming at the practical use in phase 2, we examined an application to FBR research and development of cost benefit analysis method used for the policy evaluation. Furthermore, long-term nuclear material mass flow was analyzed and the scenario of 'FBR application for the hydrogen production' is proposed, considering how FBR would be utilized for the 21st century. And, a database including the various documents and data used for evaluation was constructed. (author)

  5. Extended stability of intravenous 0.9% sodium chloride solution after prolonged heating or cooling.

    Science.gov (United States)

    Puertos, Enrique

    2014-03-01

    The primary objective of this study was to evaluate the stability and sterility of an intravenous 0.9% sodium chloride solution that had been cooled or heated for an extended period of time. Fifteen sterile 1 L bags of 0.9% sodium chloride solution were randomly selected for this experiment. Five bags were refrigerated at an average temperature of 5.2°C, 5 bags were heated at an average temperature of 39.2°C, and 5 bags were stored at an average room temperature of 21.8°C to serve as controls. All samples were protected from light and stored for a period of 199 days prior to being assayed and analyzed for microbial and fungal growth. There was no clinically significant difference in the mean sodium values between the refrigerated samples, the heated samples, and the control group. There were no signs of microbial or fungal growth for the duration of the study. A sterile intravenous solution of 0.9% sodium chloride that was heated or cooled remained stable and showed no signs of microbial or fungal growth for a period of 199 days. This finding will allow hospitals and emergency medical technicians to significantly extend the expiration date assigned to these fluids and therefore obviate the need to change out these fluids every 28 days as recommended by the manufacturer.

  6. Research and development of FBR fuel reprocessing in PNC

    International Nuclear Information System (INIS)

    Hoshino, T.

    1976-05-01

    The research program of the PNC for FBR fuel reprocessing in Japan is discussed. The general characteristics of FBR fuel reprocessing are pointed out and a comparison with LWR fuel is made. The R and D program is based on reprocessing using the aqueous Purex process. So far, some preliminary steps of the research program have been carried out, these include solvent extraction test, off-gas treatment test, voloxidation process study, solidification test of high-level liquid waste, and study of the dissolution behaviour of irradiated mixed oxide fuel. By the end of the 1980s, a pilot plant for FBR fuel reprocessing will be completed. For the design of the pilot plant, further research will be carried out in the following fields: head-end techniques; voloxidation process; dissolution and extraction techniques; waste treatment techniques. A time schedule for the different steps of the program is included

  7. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    Energy Technology Data Exchange (ETDEWEB)

    Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  8. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

    2001-01-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  9. Sintered-to-size FBR fuel

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1984-04-01

    Fabrication of sintered-to-size PuO 2 -UO 2 fuel pellets was completed for testing of proposed FBR product specifications. Approximately 6000 pellets were fabricated to two nominal diameters and two densities by cold pressing and sintering to size. Process control and correlation between test and production batches are discussed

  10. Preliminary Design of Compressor Impeller for innovative Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jekyoung; Cho, Seongkuk; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Cha, Jae Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For nuclear power plant application, applying S-CO{sub 2} Brayton cycle to Sodium cooled Fast Reactors and Small Modular Reactors are currently considered and active research is being performed by various research institutions and universities. As a part of research activities on the SCO{sub 2} Brayton cycle development for a nuclear power system, KAIST joint research team is currently working on an innovative Sodium cooled Fast Reactor (iSFR) development which utilizes S-CO{sub 2} Brayton cycle as its power conversion system. Various research subjects including reactor physics, thermo-hydraulics, material, cycle analysis and system integration are being considered as research issues currently. However, technical issues rising from dramatic change of thermodynamic property of CO{sub 2} near the critical point still remain as problems to be solved. As a result, 3D impeller model generation based on 1D mean stream line analysis results was successfully performed for non-airfoil blades. Since 3D model generation module works successfully, KAIST{sub T}MD can support 3D CFD analysis for internal flow structure in the designed impeller. Compressor loss mechanisms are complex phenomena and these are difficulties to be modeled while considering each loss mechanism separately.

  11. Conceptual core design study for Japan sodium-cooled fast reactor: Review of sodium void reactivity worth evaluation

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2012-01-01

    The conceptual core design study for a large-scale Japan sodium-cooled fast reactor (JSFR) have been carried out in the framework of the FaCT project. The reference “High-internal conversion” core can satisfy the requirements for enhanced safety, as well as achieving economic competitiveness. In order to increase the design reliability, more rigorous uncertainty evaluation is important. Development of the verification and validation methodology of the core neutronic design method is currently underway. (author)

  12. A fast track approach to commercializing the sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Boardman, C.E.; Hui, Marvin; Carroll, Douglas

    1999-01-01

    As a result of more than 50 years of Liquid Metal Reactor design and development work the basic technology is well understood. However, commercialization of the Fast Breeder Reactor (FBR) has been delayed while various approaches to achieving competitive plant and fuel cycle costs are explored, developed, and demonstrated in prototype systems. Most designers have elected to take advantage of the economy of scale but are burdened by the cost and risk associated with the need for incremental scale up through the design, construction, and operation of multiple demonstration plants. An alternative commercialization path developed by GE would utilize a modular plant design to reduce the plant construction, R and D, and economic risk associated with the need to build multiple demonstration plants to reach a competitive size'. The key question is can a modular FBR compete with alternative electrical generation systems? Recently completed studies indicate that the answer to this question is yes if the modular plant designers keep the design simple by incorporating passive safety features and optimizing the manner in which supporting service systems are shared. (author)

  13. Towards the Characterization of the Bubble Presence in Liquid Sodium of Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Cavaro, M.; Jeannot, J.P.; Payan, C.

    2013-06-01

    In a Sodium cooled Fast Reactors (SFR), different phenomena such as gas entrainment or nucleation can lead to gaseous micro-bubbles presence in the liquid sodium of the primary vessel. Although this free gas presence has no direct impact on the core neutronics, the French Atomic Energy and Alternative Energies Commission (CEA) currently works on its characterization to, among others, check the absence of risk of large gas pocket formation and to assess the induced modifications of the sodium acoustic properties. The main objective is to evaluate the void fraction values (volume fraction of free gas) and the radii histogram of the bubbles present in liquid sodium. Acoustics and electromagnetic techniques are currently developed at CEA: - The low-frequency speed of sound measurement, which allows us to link - thanks to Wood's model - the measured speed of sound to the actual void fraction. - The nonlinear mixing of two frequencies, based on the nonlinear resonance behavior of a bubble. This technique allows knowing the radius histogram associated to a bubble cloud. Two different mixing techniques are presented in this paper: the mixing of two high frequencies and the mixing of a high and a low frequency. - The Eddy-current flowmeter (ECFM), the output signal of which is perturbed by free gas presence and in consequence allows detecting bubbles. For each technique, initial results are presented. Some of them are really promising. So far, acoustic experiments have been led with an air-water experimental set-up. Micro-bubbles clouds are generated with a dissolved air flotation device and monitored by an optical device which provides reference measurements. Generated bubbles have radii range from few micrometers to several tens of micrometers. Present and future air/water experiments are presented. Furthermore, a development plan of in-sodium tests is presented in terms of a device set-up, instrumentation, modeling tools and experiments. (authors)

  14. Development of advanced methodology for defect assessment in FBR power plants

    International Nuclear Information System (INIS)

    Meshii, Toshiyuki; Asayama, Tai

    2001-03-01

    As a preparation for developing a code for FBR post construction code, (a) JSME Code NA1-2000 was reviewed on the standpoint of applying it to FBR power plants and the necessary methodologies for defect assessment for FBR plants were pointed out (b) large capacity-high speed fatigue crack propagation (FCP) testing system was developed and some data were acquired to evaluate the FCP characteristics under thermal stresses. Results showed that the extended research on the following items are necessary for developing FBR post construction code. (1) Development of assessment for multiple defects due to creep damage. Multiple defects due to creep damage are not considered in the existing code, which is established for nuclear power plants in service under negligible-creep temperature. Therefore method to assess the integrity of these multiple defects due to creep damage is necessary. (2) FCP resistance for small load. Since components of FBR power plants are designed to minimize thermal stresses, the accuracy of FCP resistance for small load is important to estimate the crack propagation under thermal stresses accurately. However, there is not a sufficient necessary FCP data for small loads, maybe because the data is time consuming. Therefore we developed a large capacity-high speed FCP testing system, made a guideline for accelerated test and acquired some data to meet the needs. Continuous efforts to accumulate small load FCP data for various materials are necessary. (author)

  15. Fast breeder reactor safety : a perspective

    International Nuclear Information System (INIS)

    Kale, R.D.

    1992-01-01

    Taking into consideration India's limited reserves of natural and vast reserves of thorium, the fast reactor route holds a great promise for India's energy supply in future. The fast reactor fueled with 239 Pu/ 238 U (unused or depleted) produces (breeds) more fissionable fuel material 239 Pu than it consumes. Calculations show that a fast breeder reactor (FBR) increases energy potential of natural uranium by about 60 times. As the fast reactor can also convert 232 Th into 233 U which is a fissionable material, it can make India's thorium reserves a source of almost inexhaustible energy supply for a long time to come. Significant advantage of FBR plants cooled by sodium and their world-wide operating experience are reviewed. There are two main safety issues of FBR, one nuclear and the other non-nuclear. The nuclear issue concerns core disruptive accident and the non-nuclear one concerns the high chemical energy potential of sodium. These two issues are analysed and it is pointed that they are manageable by current design, construction and operational practices. Main findings of safety research during the last six to eight years in West European Countries and United States of America (US) are summarised. Three stage engineered safety provision incorporated into the design of the sodium cooled Fast Breeder Test Reactor (FBTR) commissioned at Kalpakkam are explained. The important design safety features of FBTR such as primary system containment, emergency core cooling, plant protection system, inherent safety features achieved through reactivity coefficients, and natural convection cooling are discussed. Theoretical analysis and experimental research in fast reactor safety carried out at the Indira Gandhi Centre for Atomic Research during the past some years are reviewed. (M.G.B.)

  16. International standardization of safety requirements for fast reactors

    International Nuclear Information System (INIS)

    2011-06-01

    Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

  17. Identification of important phenomena under sodium fire accidents based on PIRT process with factor analysis in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki

    2016-01-01

    The PIRT (Phenomena Identification and Ranking Table) process is an effective method to identify key phenomena involved in safety issues in nuclear power plants. The present PIRT process is aimed to validate sodium fire analysis codes. Because a sodium fire accident in sodium-cooled fast reactor (SFR) involves complex phenomena, various figures of merit (FOMs) could exist in this PIRT process. In addition, importance evaluation of phenomena for each FOM should be implemented in an objective manner under the PIRT process. This paper describes the methodology for specification of FOMs, identification of associated phenomena and importance evaluation of each associated phenomenon in order to complete a ranking table of important phenomena involved in a sodium fire accident in an SFR. The FOMs were specified through factor analysis in this PIRT process. Physical parameters to be quantified by a sodium fire analysis code were identified by considering concerns resulting from sodium fire in the factor analysis. Associated phenomena were identified through the element- and sequence-based phenomena analyses as is often conducted in PIRT processes. Importance of each associated phenomenon was evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses. (author)

  18. In service inspection and repair of sodium cooled ASTRID prototype

    Energy Technology Data Exchange (ETDEWEB)

    Baque, F.; Jadot, F. [French Atomic Commission, Cadarache Centre, 13108 Saint Paul lez Durance Cedex, (France); Marlier, R. [AREVA, 10 rue Recamier, 69456 Lyon cedex 06, (France); Saillant, J-F. [AREVA/NDE Solutions, 4 rue Thomas Dumorey, BP 70385, 71109 Chalon sur Saone Cedex, (France); Delalande, V. [EDF R and D, 6, quai Watier, 78400 Chatou, (France)

    2015-07-01

    In the frame of the large R and D work which is performed for the future ASTRID sodium cooled prototype, In Service Inspection and Repair (ISI and R) has been identified as a major issue to be taken into account in order to enlarge the plant safety, to consolidate its availability and to protect the associated investment. After the first part of pre-conceptual design phase (2008-2012), the running second part of pre-conceptual phase (2013-2015) allows to increase the ISI and R tool ability for immersed sodium structures of ASTRID, at about 200 deg. C, on the basis of consolidated specifications and thanks to their qualification through more and more realistic laboratory tests and simulation with CIVA code. ISI and R items are being developed and qualified during a pluri-annual program which mainly deals with the reactor block structures, the primary components and circuit, and the Power Conversion System. It ensures a strong connection between the reactor designers and inspection specialists, as the optimization of inspectability and repairability is looked at: this already induced specific rules for design, in order to shorten and ease the ISI and R operations, which have been merged into RCC-MRx rules. In the frame of increasing technology readiness level with corresponding performance demonstration, this paper presents R and D dealing with the ISI and R items: it highlights the sensor development (both ultrasonic and electromagnetic concepts, compatible with sodium at 200 deg. C), then their applications for ASTRID structure control (under sodium telemetry, imaging and NDE). Activity for repair is also presented (a single laser tool for sodium sweeping, machining and welding), and finally the effort for associated robotic (generic program for ASTRID applications, specific technological tools for sodium medium, tight immersed bell). The main results of testing and simulation are given for telemetry, vision, NDE applications, laser process repair and under sodium

  19. Heat-transfer in a partially-blocked sodium-cooled rod bundle

    International Nuclear Information System (INIS)

    Han, J.T.

    1979-01-01

    Heat transfer coefficients were experimentally determined for 31-rod sodium-cooled bundle with a 6-subchannel central blockage. The Nusselt number is presented as a function of the Peclet number for both the free flow region undisturbed by the blockage and the wake region immediately downstream of the blockage. Results are compared with the existing correlations for liquid metals. The heat transfer coefficient was generally higher in the unblocked free flow region than in the wake region. A leak at the blockage improved the heat transfer coefficient in the wake region

  20. Development of sodium disposal technology. Experiment of sodium compound solidification process

    International Nuclear Information System (INIS)

    Matsumoto, Toshiyuki; Ohura, Masato; Yatoh, Yasuo

    2007-07-01

    A large amount of sodium containing radioactive waste will come up at the time of final shutdown/decommission of FBR plant. The radioactive waste is managed as solid state material in a closed can in Japan. As for the sodium, there is no established method to convert the radioactive sodium to solid waste. Further, the sodium is highly reactive. Thus, it is recommended to convert the sodium to a stable substance before the solidification process. One of the stabilizing methods is conversion of sodium into sodium hydroxide solution. These stabilization and solidification processes should be safe, economical, and efficient. In order to develop such sodium disposal technology, nonradioactive sodium was used and a basic experiment was performed. Waste-fluid Slag Solidification method was employed as the solidification process of sodium hydroxide solution. Experimental parameters were mixing ratio of the sodium hydroxide and the slag solidification material, temperature and concentration of the sodium hydroxide. The best parameters were obtained to achieve the maximum filling ratio of the sodium hydroxide under a condition of enough high compressive strength of the solidified waste. In a beaker level test, the solidified waste was kept in a long term and it was shown that there was no change of appearance, density, and also the compressive strength was kept at a target value. In a real scale test, homogeneous profiles of the density and the compressive strength were obtained. The compressive strength was higher than the target value. It was shown that the Waste-fluid Slag Solidification method can be applied to the solidification process of the sodium hydroxide solution, which was produced by the stabilization process. (author)

  1. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  2. Operating Temperatures of a Sodium-Cooled Exhaust Valve as Measured by a Thermocouple

    Science.gov (United States)

    Sanders, J. C.; Wilsted, H. D.; Mulcahy, B. A.

    1943-01-01

    A thermocouple was installed in the crown of a sodium-cooled exhaust valve. The valve was then tested in an air-cooled engine cylinder and valve temperatures under various engine operating conditions were determined. A temperature of 1337 F was observed at a fuel-air ratio of 0.064, a brake mean effective pressure of 179 pounds per square inch, and an engine speed of 2000 rpm. Fuel-air ratio was found to have a large influence on valve temperature, but cooling-air pressure and variation in spark advance had little effect. An increase in engine power by change of speed or mean effective pressure increased the valve temperature. It was found that the temperature of the rear spark-plug bushing was not a satisfactory indication of the temperature of the exhaust valve.

  3. The Modification of Sodium Polyacrylate Water Solution Cooling Properties by AL2O3

    Directory of Open Access Journals (Sweden)

    Wojciech Gęstwa

    2010-01-01

    Based on cooling curves, it can be concluded that for the water solution of sodium polyacrylate with AL2O3 nanoparticles in comparison to water and 10% polymer water solution lower cooling speed is obtained. The cooling medium containing nanoparticles provides lower cooling speed in the smallest surface austenite occurance (500–600 C in the charts of the CTP for most nonalloy structural steels and low-alloy steels. However lower cooling temperature at the beginning of martensitic transformation causes the formation of smaller internal stresses, leading to smaller dimensional changes and hardening deformation. For the quenching media the wetting angle was appointed by the drop-shape method. These studies showed the best wettability of polymer water solution (sodium polyacrylate with the addition of AL2O3 nanoparticles, whose wetting angle was about 65 degrees. Obtaining the smallest wetting angle for the medium containing nanoparticles suggests that the heat transfer to the cooling medium is larger. This allows slower cooling at the same time ensuring its homogeneity. The obtained values of wetting angle confirm the conclusions drawn on the basis of cooling curves and allowus to conclude that in the case of the heat transfer rate it will have a lower value than for water and 10% polymer water solution. In the research on hardened carburized steel samples C10 and 16MnCr5 surface hardness, impact strength and changes in the size of cracks in Navy C-ring sample are examined. On this basis of the obtained results it can be concluded that polymer water solution with nanoparticles allows to obtain a better impact strength at comparable hardness on the surface. Research on the dimensional changes on the basis of the sample of Navy C-ring also shows small dimensional changes for samples carburized and hardened in 10% polymer water solution with the addition of nanoparticles AL2O3. Smaller dimensional changes were obtained for samples of steel 16MnCr5 thanfar C10. The

  4. Specialists' meeting on bellows for sodium systems. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-02-01

    The Specialists' Meeting on Bellows for Sodium Systems was sponsored by the IAEA on the recommendation of the International Working Group on Fast Reactors (IWGFR) was attended by participants and observers from France, the Federal Republic of Germany, Italy, Japan, the Netherlands, the United Kingdom and the IAEA. The purpose of the meeting was to provide forum for exchanging views on application of bellows for FBR use, problems found in service in sodium systems, design and fabrication of bellows for sodium systems and studies necessary for estimation and improvement of reliability of bellows in long term use under the condition of high temperature sodium. The technical parts of the meeting were divided into five major sessions, as follows: Experience of Bellows Applications for Sodium Systems; Design and Analysis; Fabrication; In-Service Inspection and Repair; Research Work.

  5. Specialists' meeting on bellows for sodium systems. Summary report

    International Nuclear Information System (INIS)

    1980-02-01

    The Specialists' Meeting on Bellows for Sodium Systems was sponsored by the IAEA on the recommendation of the International Working Group on Fast Reactors (IWGFR) was attended by participants and observers from France, the Federal Republic of Germany, Italy, Japan, the Netherlands, the United Kingdom and the IAEA. The purpose of the meeting was to provide forum for exchanging views on application of bellows for FBR use, problems found in service in sodium systems, design and fabrication of bellows for sodium systems and studies necessary for estimation and improvement of reliability of bellows in long term use under the condition of high temperature sodium. The technical parts of the meeting were divided into five major sessions, as follows: Experience of Bellows Applications for Sodium Systems; Design and Analysis; Fabrication; In-Service Inspection and Repair; Research Work

  6. Experience with, and programme of, FBR and HWR development in Japan

    International Nuclear Information System (INIS)

    Iida, M.; Sawai, S.; Nomoto, S.

    1983-01-01

    Nuclear power generation in Japan is moving forward on the long-term development programme of nuclear power from the LWR to the FBR, essentially in the same way as in other advanced nuclear countries. In this development programme the unique HWR is also included; it can use plutonium produced in LWRs together with depleted uranium before the introduction of commercial FBRs. This report describes the status of the FBR and HWR development project being carried out by the Power Reactor and Nuclear Fuel Development Corporation (PNC) based upon the Long-Term Programme on Research, Development and Utilization of Nuclear Energy in Japan. Operational experience and technical results are shown for the experimental fast reactor JOYO (100 MW(th)), which reached initial criticality in 1977. The status of the 280 MW(e) prototype reactor MONJU, under construction as of 1982, is described. The conceptual design of the subsequent 1000 MW(e) demonstration plant is outlined, as is additional future planning. Research and development results, mainly carried out at Oarai Engineering Center of PNC, are shown. The 165 MW(e) prototype FUGEN is a heavy-water-moderated, boiling-light-water-cooled, pressure-tube-type reactor which uses plutonium mixed-oxide fuel. This report describes the relationship of the fuel cycle to the HWR in Japan and also discusses the operational experience of the prototype FUGEN, which has operated since 1979. Also described is the design of the 600 MW(e) demonstration plant and the programme of related research and development. (author)

  7. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

  8. Preparation of a monoenergetic sodium beam by laser cooling and deflection

    International Nuclear Information System (INIS)

    Nellessen, J.; Sengstock, K.; Muller, J.H.; Ertmer, W.; Wallis, H.

    1989-01-01

    This paper reports on a sodium atomic beam with a density of approx. 10 5 at cm 3 within a velocity interval of less than 3 m/s with a mean velocity of typically 50-160 m/s which has been produced by laser deflection of a laser cooled atomic beam. Laser cooling with the frequency chirp method decelerates and cools a considerable part of an atomic beam into a narrow velocity group with a temperature of approx 30 mK as a part of the resulting atomic beam. This velocity group has been selectively deflected up to 30 degrees - 40 degrees using a light field with k vectors always perpendicular to the atomic trajectory. If the light field is prepared by use of a cylindrical lens, the angle of deflection is nearly independent from the actual orbit radius. For a laser frequency detuning of about one natural linewidth to the red, the strong frequency dependence of the light pressure force leads to a beam collimation via detuning-locking of the atomic trajectory. To avoid optical pumping we used a frequency modulated laser beam with a sideband spacing matched to the hyperfine splitting of the ground state. As the cooling was performed by the frequency chirp method, one can use a part of the cooling laser beam as deflecting laser beam. Typical velocity distributions in the deflected and undeflected atomic beam, measured 22 cm downstream the deflection zone. It shows the perfect transfer of the cooled velocity group from the laser cooled beam into the deflected beam; curve c) shows as comparison the result for the deflection of the initial thermal atomic beam

  9. Continuous Ethanol Production Using Immobilized-Cell/Enzyme Biocatalysts in Fluidized-Bed Bioreactor (FBR)

    Energy Technology Data Exchange (ETDEWEB)

    Nghiem, NP

    2003-11-16

    The immobilized-cell fluidized-bed bioreactor (FBR) was developed at Oak Ridge National Laboratory (ORNL). Previous studies at ORNL using immobilized Zymomonas mobilis in FBR at both laboratory and demonstration scale (4-in-ID by 20-ft-tall) have shown that the system was more than 50 times as productive as industrial benchmarks (batch and fed-batch free cell fermentations for ethanol production from glucose). Economic analysis showed that a continuous process employing the FBR technology to produce ethanol from corn-derived glucose would offer savings of three to six cents per gallon of ethanol compared to a typical batch process. The application of the FBR technology for ethanol production was extended to investigate more complex feedstocks, which included starch and lignocellulosic-derived mixed sugars. Economic analysis and mathematical modeling of the reactor were included in the investigation. This report summarizes the results of these extensive studies.

  10. The questions of liquid metal two-phase flow modelling in the FBR core channels

    International Nuclear Information System (INIS)

    Martsiniouk, D.Ye.; Sorokin, A.P.

    2000-01-01

    The two-fluid model representation for calculations of two-phase flow characteristics in the FBR fuel pin bundles with liquid metal cooling is presented and analysed. Two conservation equations systems of the mass, momentum and energy have been written for each phase. Components accounted the mass-, momentum- and heat transfer throughout the interface occur in the macro-field equations after the averaging procedure realisation. The pattern map and correlations for two-fluid model in vertical liquid metal flows are presented. The description of processes interphase mass- and heat exchange and interphase friction is determined by the two-phase flow regime. The opportunity of the liquid metal two-phase flow regime definition is analysed. (author)

  11. FBR type reactors

    International Nuclear Information System (INIS)

    Otsuka, Masaya; Yamakawa, Masanori; Goto, Tadashi; Ikeuchi, Toshiaki; Yamaki, Hideo.

    1986-01-01

    Purpose: To prevent thermal deformation and making the container compact by improving the cooling performance of main container walls. Constitution: A pipeway is extended from a high pressure plenum below the reactor core and connected to the lower side of the flow channel at the inside of a thermal shielding layer disposed to the inside of the main container wall. Low pressure sodium sent from the low temperature plenum into the high pressure plenum is introduced to the pipeway, caused to uprise in the inside flow channel, then turned for the direction, caused to descend in the outer side flow channel between the main container and the inside flow channel and then returned to the low temperature plenum. A heat insulating layer disposed with argon gas is installed to the inside of the flow channel to reduce the temperature change applied upon reactor scram. An annular linear induction pump capable of changing the voltage polarity is disposed at the midway of the pipeway and the polarity is switched such that the direction of flow of the liquid sodium is exerted as a braking force upon rated operation, whereas exerted as a pumping force upon reactor scram. (Sekiya, K.)

  12. Tests of the heat transfer characteristic of air cooler during cooling by natural convection of the Fast Breeder Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purpose of this study is to confirm the heat transfer characteristics of the air cooler (AC) of the Fast Breeder Reactor(FBR) which has a function to remove the residual heat of the reactor by heat exchange between sodium and air in natural convection region if electric power would be lost. In order to confirm the characteristics of the AC installed in the FBR plant, the heat transfer test by using the AC which is installed in the sodium test loop owned by Toshiba Corporation has been planned. In this study, the heat transfer characteristic tests were performed by using the AC in sodium test loop, and the CFD analyses were conducted to evaluate the test results and the heat transfer characteristics of the plant scale AC at the condition of natural convection. In addition, the elemental tests to confirm the influence of the heat transfer tube placement by using the heat transfer tube of the same specification as the AC of Monju were performed. (author)

  13. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    Energy Technology Data Exchange (ETDEWEB)

    Scaller, K; Vrillon, B

    1980-02-01

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component.

  14. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    International Nuclear Information System (INIS)

    Scaller, K.; Vrillon, B.

    1980-01-01

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

  15. Sodium leak detection system for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Modarres, D.

    1991-01-01

    This patent describes a device for detecting sodium leaks from a reactor vessel of a liquid sodium cooled nuclear reactor the reactor vessel being concentrically surrounded by a a containment vessel so as to define an airtight gap containing argon. It comprises: a light source for generating a first light beam, the first light beam having first and second predominant wavelengths, the first wavelength being substantially equal to an absorption line of sodium and the second wavelength being chosen such that it is not absorbed by sodium and argon; an optical multiplexer optically coupled to the light source; optically coupled to the multiplexer, each of the sensors being embedded in the containment vessel of the reactor, each of the sensors projecting the first light beam into the gap and collecting the first light beam after it has reflected off of a surface of the reactor vessel; a beam splitter optically coupled to each of the sensors through the multiplexer, the beam splitter splitting the first light beam into second and third light beams of substantially equal intensities; a first filter dispersed within a path of second light beam for filtering the second wavelength out of the third light beam; first and second detector beams disposed with in the paths of the second and third light beams so as to detect the intensities of the second and third light beams, respectively; and processing means connected to the first and second detector means for calculating the amount of the first wavelength which is absorbed when passing through the argon

  16. Development of computer code on sodium-water reaction products transport

    International Nuclear Information System (INIS)

    Arikawa, H.; Yoshioka, N.; Suemori, M.; Nishida, K.

    1988-01-01

    The LMFBR concept eliminating the secondary sodium system has been considered to be one of the most promissing concepts for offering cost reductions. In this reactor concept, the evaluation of effects on reactor core by the sodium-water reaction products (SWRPs) during sodium-water reaction at primary steam generator becomes one of the major safety issues. In this study, the calculation code was developed as the first step of the processes of establishing the evaluation method for SWRP effects. The calculation code, called SPROUT, simulates the SWRPs transport and distribution in primary sodium system using the system geometry, thermal hydraulic data and sodium-water reacting conditions as input. This code principally models SWRPs behavior. The paper contain the modelings for SWRPs behaviors, with solution, precipation, deposition and so on, and the results and discussions of the demonstration calculation for a typical FBR plant eliminating the secondary sodium system

  17. KNK II, Compact Sodium-Cooled Reactor in the Nuclear Research Center Karlsruhe

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives an overview of the project of the sodium-cooled fast reactor KNK II in the nuclear research center KfK in Karlsruhe. This test reactor was the preparatory stage of the prototype plant SNR 300 and had several goals: to train operating personal, to practice the licensing procedures in Germany, to get experience with the sodium technology and to serve as a test bed for fast breeder core components. The report contains contributions of KfK as the owner and project managing organization, of INTERATOM as the design and construction company and of the KBG as the plant operating organization. Experience with and results of relevant aspects of the project are tackled: project management, reactor core and component design, safety questions and licensing, plant design and test programs [de

  18. Numerical study of the underexpanded nitrogen jets submerged into liquid sodium in the frame of Sodium-cooled Fast Reactor (SFRs)

    International Nuclear Information System (INIS)

    Chen, F.; Allou, A.; Parisse, J.D.

    2017-01-01

    The study of the consequences of a gas leakage in the secondary/ tertiary heat exchangers is one of the essential points in the safety analysis of Sodium-cooled Fast nuclear Reactors (SFRs). This work is in the frame of the technology of the Compact plates Sodium-Gas heat Exchangers (ECSG) which is an alternative to conventional steam Rankine cycles. The overpressure of the tertiary nitrogen loop causes the formation of underexpanded gas jets submerged in the liquid sodium. In order to establish a safety evaluation, it would be an asset to be able to estimate the leakage. The gas leak detection by the acoustic method based on the bubbles field has been proposed. It requires then a delicate knowledge of the bubble field. This work contributes to development a numerical tool and its validation to model the transport and the production of bubbles in the downstream of underexpanded gas jets. The code CANOP modeling bi-phasic compressible flow is investigated under the actual condition of the underexpanded nitrogen jets submerged in the liquid sodium in an ECSG channel. Expensive computational cost is limited by using an Adaptive Mesh Refinement. (author)

  19. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  20. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

    The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

  1. Long-term logistic analysis of FBR introduction strategy: avoiding both uranium and plutonium shortage

    International Nuclear Information System (INIS)

    Suzuki, T.

    1995-01-01

    Despite comfortable predictions on short to mid-term uranium resources, there is still a concern about long-term availability of competitive uranium resources. In order to achieve substantial uranium saving, early introduction of Fast Breeder Reactor (FBR) is desirable. But it is also known that rapid introduction of FBR could result in plutonium storage. Will there be enough plutonium on a global scale to sustain fast FBR growth? is there any other way to save uranium resource? This paper concludes that multi-option strategies to achieve flexible long-term strategy to avoid both uranium and plutonium storage are desirable. (authors)

  2. Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Jang, Sung Hyun; Takata, Takashi; Yamaguchi, Akira; Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-01-01

    Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

  3. Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Sung Hyun; Takata, Takashi [Graduate School of Engineering, Osaka University, Osaka (Japan); Yamaguchi, Akira [Graduate School of Engineering, The University of Tokyo, Ibaraki (Japan); Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki [Japan Atomic Energy Agency, Ibaraki (Japan)

    2015-10-15

    Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

  4. Evaluation of tube rupture simulation test (TRUST-1) for FBR steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Hayashida, Yoshihiko; Hamada, Hirotsugu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-06-01

    The intermediate water leak in an FBR Steam Generator (SG) causes a high temperature and corrosive sodium-water reaction jet. In such cases, it is necessary to evaluate the wastage and overheating rupture behavior of heat transfer tubes. Especially, in the large SG that aims at high temperature of sodium and high temperature/pressure of water, the establishment of the rational evaluation method is important. In this paper, as a basic experiment to make clear the phenomenon of overheating rupture, tests and analysis of Tube Rupture Simulation Test-1 (TRUST-1) were conducted. TRUST-1 simulates the overheating rupture of the tube made of Mod.9Cr-1Mo steel by nitrogen gas pressurization and quick induction heating. The result of TRUST-1 are as follows: (1) The breaking strength predicted by the internal pressure is larger than the tensile strength of the tube material. (2) The margin of the breaking strength from the tensile strength of the tube material has a tendency of decreasing with the heating rate, especially in the lower temperature region. (3) Using an theoretical formula that is deduced from the steady creep model and appropriate experimental coefficients that are determined by the test data, the breaking strength can be reasonably evaluated. (author)

  5. Water flow simulation of the flow-induced vibration phenomenon of the thermowell in the prototype-FBR 'Monju'

    International Nuclear Information System (INIS)

    Anoda, Yoshinari; Kuroda, Takeshi; Kondo, Masaya; Murata, Hideo

    1996-06-01

    On December 8, 1995 a sodium leak event occurred in the secondary heat transport system (SHTS) of the prototype fast breeder reactor (FBR), Monju, owned and operated by the Power Reactor and Nuclear Fuel Development Corporation (PNC). The direct cause of the leak was a break of a thermowell installed in the loop piping of the SHTS. The break of the thermowell is now believed to have resulted from the flow-induced vibrations due to vortex shedding from the thermowell subjected to a crossflow of sodium. The Japan Atomic Energy Research Institute has conducted a series of water flow model experiments on the flow-induced vibrations of the thermowell to contribute to the post-factor analyses of the event conducted by the Investigation Taskforce on the Sodium Leak Accident in Monju which was established by the Science and Technology Agency (STA) after this event. The experiments were performed for a wide range of experimental conditions including the condition corresponding to the operating condition of the Monju's thermowell and showed the relationship between the vortex shedding pattern and the vibration mode as well as influence of the damping (stability) parameter on the amplitude of vibration. (author)

  6. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  7. Materials Performance in Sodium-Cooled Fast Reactors: Past, Present, and Future

    International Nuclear Information System (INIS)

    Natesan, K.; Li Meimei

    2013-01-01

    • This paper gives an overview of the requirements, selection, and performance of materials for in-core and out-of-core components in SFRs. • Globally, sodium-cooled fast reactors have been designed, built, and operated in several countries. A substantial database exists for the existing materials on their functional and mechanical performance. • The 60-yr design life of the SFR presents a significant challenge to the development of database, extrapolation/prediction of long-term performance, and high-temperature design methodology for the structural components. • Licensing of SFR requires a valid assessment of the environmental effects (irradiation, thermal aging, and sodium) on materials performance. • Advanced materials such as, ODS alloys for cladding, Gr91 and 92 F/M steels, and austenitic alloys such as NF709 for structures can improve the economy, safety, and flexibility of SFRs. A substantial database is needed for all these materials and global effort is underway to develop the needed information through experimentation and modeling

  8. Advanced hybrid process with solvent extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR

    International Nuclear Information System (INIS)

    Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo

    2008-01-01

    Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)

  9. Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

    Science.gov (United States)

    Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

    2018-01-01

    ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical

  10. Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

    OpenAIRE

    Verma, Vasudha

    2017-01-01

    Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

  11. Fuel supply demand balances for future FBR commercialization: impacts on plutonium pricing and reactor design

    International Nuclear Information System (INIS)

    Braun, C.; Zebroski, E.L.

    1985-01-01

    Plutonium supply and demand balances for fast breeder reactor (FBR) commercialization post-2000 were computed to determine: (a) the maximum supportable number of FBRs that could be installed based on plutonium availability considerations and (b) the feasibility of a reasonable FBR capacity growth case assuming slow introduction post-2010 and rapid capacity growth post-2035. The purpose of the analysis was to determine the outer limitation on the maximum future FBR introduction, or the bounds of a possible plutonium-limited introduction rate, and to estimate the reasonableness of a more limited capacity growth case

  12. Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1978-10-01

    The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

  13. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  14. Reaction velocity of sodium hydration in humid air and sodium carbonation in humid carbon dioxide atmosphere. Fundamental study on sodium carbonate process in FBR bulk sodium coolant disposal technology

    International Nuclear Information System (INIS)

    Tadokoro, Yutaka; Yoshida, Eiichi

    1999-11-01

    A sodium carbonate processing method, which changes sodium to sodium carbonate and/or sodium bicarbonate by humid carbon dioxide, has been examined and about to be applied to large test loops dismantling. However, that the basic data regarding the progress of the reaction is insufficient on the other hand, is a present condition. The present report therefore aims at presenting basic data regarding the reaction velocity of sodium hydration in humid air and sodium carbonation in humid carbon dioxide atmosphere, and observing the reaction progress, for the application to large test loops dismantling. The test result is summarized as follows. (1) Although the reaction velocity of sodium varied with sodium specimen sizes and velocity measurement methods, the reaction velocity of sodium hydration was in about 0.16 ∼ 0.34 mmh -1 (0.016 ∼ 0.033g cm -2 h -1 , 6.8x10 -4 ∼ 1.4x10 -3 mol cm -2 h -1 ) and that of sodium carbonation was in about 0.16 ∼ 0.27mmh -1 (0.016 ∼ 0.023g cm -2 h -1 , 6.8x10 -4 ∼ 1.1x10 -3 mol cm -2 h -1 ) (26 ∼ 31degC, RH 100%). (2) The reaction velocity of sodium in carbon dioxide atmosphere was greatly affected by vapor partial pressure (absolutely humidity). And the velocity was estimated in 0.08 ∼ 0.12mmh -1 (0.008 ∼ 0.012g cm -2 h -1 , 3.4x10 -4 ∼ 5.2x10 -4 mol cm -2 h -1 ) in the carbon dioxide atmosphere, whose temperature of 20degC and relative humidity of 80% are assumed real sodium carbonate process condition. (3) By the X-ray diffraction method, NaOH was found in humid air reaction product. Na 2 CO 3 , NaHCO 3 were found in carbon dioxide atmosphere reaction product. It was considered that Sodium changes to NaOH, and subsequently to NaHCO 3 through Na 2 CO 3 . (4) For the application to large test loops dismantling, it is considered possible to change sodium to a target amount of sodium carbonate (or sodium bicarbonate) by setting up gas supply quantity and also processing time appropriately according to the surface area

  15. Outline of the seismic design guideline of an FBR - a tentative draft

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Ohtsubo, Hideomi; Nakamura, Hideharu; Matsuura, Shinichi; Hagiwara, Yutaka; Yuhara, Tetsuo; Hirayama, Hiroshi; Kokubo, Kunio; Ooka, Yuji.

    1993-01-01

    Central Research Institute of Electric Power Industry (Japan) is carrying out the Demonstration Test and Research Program of Buckling of FBR (FY 1987-FY 1993). The first half of the research program was finished after establishing a seismic buckling design guideline (a tentative draft). The purpose of this paper is to describe the dynamic buckling characteristics of FBR main vessels and the outline of the rationalized buckling design guideline for seismic loadings. (orig.)

  16. Contribution to perfecting eddy current testing of steam generator tubes of sodium cooled breeders: description of the Monacault loop for the study of sodium deposit influence

    International Nuclear Information System (INIS)

    Lapicore, A.; Lemarquis, J.C.; Oberlin, C.; Pigeon, M.

    1981-12-01

    In the event of sodium-water reaction in the steam generator of a sodium cooled breeder reactor, it is essential to be able to monitor the local loss of thickness of the tubes located in the reaction area. A method for monitoring the tubes by an eddy current probe is being developed for Super Phenix. The sodium deposits on the outer wall of the tubes, as well as their prolonged contact with high temperature sodium are likely to bring about a change in the signals picked up. A test loop, Monacault, has been built in order to clarify the importance of these parameters (effect of sodium deposits, reproducibility of the wetting at different temperatures). It includes three test cells containing the sample tubes having a total of 61 standard defects to be tested. The first results on the wetting of tubes are given and discussed [fr

  17. Fourteenth annual meeting of the International Working Group on Fast Reactors. Summary report. Part II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1981-05-01

    This report includes description of the state-of-the art in the field of fast reactor technology, research and development, in France, Belgium, India, Italy, USSR, USA, UK, Switzerland, and European Union. The emphasis in the majority of the reports is on the FBR safety issues, sodium cooling system, fuel elements development, reactor materials testing, risk assessment.

  18. Fourteenth annual meeting of the International Working Group on Fast Reactors. Summary report. Part II

    International Nuclear Information System (INIS)

    1981-05-01

    This report includes description of the state-of-the art in the field of fast reactor technology, research and development, in France, Belgium, India, Italy, USSR, USA, UK, Switzerland, and European Union. The emphasis in the majority of the reports is on the FBR safety issues, sodium cooling system, fuel elements development, reactor materials testing, risk assessment

  19. Control of radioactive material transport in sodium-cooled reactors

    International Nuclear Information System (INIS)

    Brehm, W.F.

    1980-03-01

    The Radioactivity Control Technology (RCT) program was established by the Department of Energy to develop and demonstrate methods to control radionuclide transport to ex-core regions of sodium-cooled reactors. This radioactive material is contained within the reactor heat transport system with any release to the environment well below limits established by regulations. However, maintenance, repair, decontamination, and disposal operations potentially expose plant workers to radiation fields arising from radionuclides transported to primary system components. This paper deals with radioactive material generated and transported during steady-state operation, which remains after 24 Na decay. Potential release of radioactivity during postulated accident conditions is not discussed. The control methods for radionuclide transport, with emphasis on new information obtained since the last Environmental Control Symposium, are described. Development of control methods is an achievable goal

  20. Method for selecting FBR development strategies in the presence of uncertainty

    International Nuclear Information System (INIS)

    Fraley, D.W.; Burnham, J.B.

    1981-12-01

    This report describes the methods used to probabilistically analyze data related to the uranium supply the FBR's competitive dates, development strategies' time and costs, and economic benefits. It also describes the econometric methods used to calculate the economic risks of mistiming the development. Seven strategies for developing the FBR are analyzed. The various measures of a strategy's performance - timing, costs, benefits, and risks - are combined into several criteria which are used to evaluate the seven strategies. Methods are described for selecting a strategy based on a number of alternative criteria

  1. Impact of nuclear data on sodium-cooled fast reactor calculations

    International Nuclear Information System (INIS)

    Aures, A.; Bostelmann, F.; Zwermann, W.; Velkov, K.

    2016-01-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors. (authors)

  2. Treatment of PAHs in waters using the GAC-FBR process

    International Nuclear Information System (INIS)

    Hickey, R.F.; Sunday, A.; Wagner, D.; Groshko, V.; Rajan, R.V.; Leuschner, A.; Hayes, T.D.

    1995-01-01

    Pilot studies were conducted to determine the utility of the granular activated carbon fluidized-bed reactor (GAC-FBR) process to treat groundwater from manufactured gas plant (MGP) sites containing polycyclic aromatic hydrocarbons (PAHs) and a process effluent water from a deep subsurface dense, nonaqueous-phase liquid (DNAPL) removal process at an MGP site. Removal of naphthalene exceeded 99.9%, and overall PAH removals of 99+% were observed at organic loading rates (OLRs) exceeding 4-kg chemical oxygen demand (COD)/m 3 -d and a hydraulic retention time (HRT) of about 6 min. Analysis of PAHs accumulated on GAC and oxygen consumption clearly demonstrated that removal of 2- to 4-ring PAHs was due primarily to biological oxidation and not to adsorption. Analysis of influent and effluent samples using Microtox reg-sign indicated removal of toxicity. Full-scale application of the GAC-FBR process has begun at a Superfund site in Pennsylvania. The GAC-FBR is being used to treat a 15-gal per min (gpm) process effluent flow from a subsurface DNAPL removal process. Initial results confirm the ability of the process to treat PAHs at high OLRs and short HRTs

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

    International Nuclear Information System (INIS)

    2003-10-01

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

  4. Neutronics optimization of LiPb-He dual-cooled fuel breeding blanket for the fusion-driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican

    2002-01-01

    The concept of the liquid Li 17 Pb 83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239 Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

  5. Characteristics of polyethylene and zirconium-hydride moderator for the NSRR tests

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yamazaki, Toshi; Sobajima, Makoto.

    1994-03-01

    Pulse irradiation tests of FBR fuels under the sodium cooling conditions are planned for the phase III program in the NSRR (Nuclear Safety Research Reactor), following the phase I and II programs of the LWR fuel tests under the simulated RIA (Reactivity Initiated Accident) conditions. A proto-type irradiation capsule for the FBR fuel rod tests and a sodium loop to purify and to charge sodium into the capsule are under construction for the tests. In the NSRR tests, neutron moderator is needed to thermalize neutrons from the driver core and to subject transient energy high enough to cause the test fuel failure. The light water has been used for the NSRR LWR fuel tests as the coolant/moderator material. Polyethylene and zirconium-hydride are candidates of the moderator for the FBR fuel tests. The capability of the moderators are investigated in the pulse irradiation tests in the NSRR. Both of the moderators indicated good capability of realizing high thermal neutron flux to subject energy depositions comparable to the light water or higher. Estimations by the SRAC code system indicated reasonable good agreement with the test results. In addition, heating tests of the moderators did not cause gas decomposition nor dissociation, indicating that the moderators are operative at temperatures up to 300degC. (author)

  6. Hydrogen transfer preventive device in FBR power plant

    International Nuclear Information System (INIS)

    Hoshi, Yuichi.

    1987-01-01

    Purpose: To prevent transfer of hydrogen, etc. in FBR power plant. Constitution: Since H 2 permeates heat conduction pipes in a steam generator, it is necessary to eliminate all of permeation hydrogen, etc. by primary cold traps particularly in the case of saving the intermediate heat exchange. In view of the above, the heat conduction pipes of the steam generator are constituted as a double pipe structure and helium gases are recycled through the gaps thereof and hydrogen traps are disposed to the recycling path. H 2 released into water flowing through the inside of the inner pipe is permeated through the inner pipe and leached into the gap, but the leached H 2 is carried by the helium recycling stream to the hydrogen trap and then the H 2 stream removed with H 2 is returned to the gaps. In this way, the capacity of the primary cold traps disposed in the liquid sodium recycling circuit can be reduced remarkably and the capacity of the purifying device, if an intermediate heat exchanger is disposed, is also reduced to decrease the plant cost. Further, diffusion of deleterious gases from the primary to the secondary circuits can be prevented as well. (Kamimura, M.)

  7. Analytical evaluation of local fault in sodium cooled small fast reactor (4S). Preliminary evaluation of partial blockage in coolant channel

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Ueda, Nobuyuki

    2007-01-01

    Local faults are fuel failures that result from heat removal imbalance within a single subassembly especially in FBRs. Although the occurrence frequency of local faults is quite low, the licensing body required local faults evaluations in previous FBR plants to confirm the potential for the occurrence of severe fuel subassembly failure and its propagation. A conceptual design of 4S (Super-Safe, Small and Simple) is a sodium cooled fast reactor, which aims at an application to dispersed energy source and long core lifetime. It has a dense arrangement of fuel pins to achieve a long lifetime. Therefore, from the viewpoint of thermal hydraulics, the 4S reactor is considered to have more potential for coolant boiling and fuel pin failure caused by formation of local blockage, comparing these potential in the conventional FBRs. The objective of the present study is to evaluate the effect of local blockage on the coolant flow pattern and temperature rise in the 4S-type fuel subassembly under the normal operation condition. A series of three-dimensional thermal-hydraulic analysis in a single subassembly with local blockage was conducted by the commercialized CFD code 'PHOENICS'. Analytical results show that the peak coolant temperature behind the blockage rises with increasing the blockage area, however, the coolant boiling does not occur under the present analytical conditions. On the other hand, it is found that the liquid phase formation caused by eutectic reactions will occur between the metallic fuel and the cladding under the local blockage condition. However, the penetration rate of liquid phase at fuel-cladding interface is quit low. Therefore, it is expected that rapid fuel pin failure and its propagation to surrounding pins due to liquid phase formation will not occur. (author)

  8. Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Masato Ando; Shigenobu Kubo; Yoshio Kamishima; Toru Iitsuka

    2006-01-01

    The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)

  9. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  10. Defect analysis of BaSrFBr:Eu irradiated by X-ray

    International Nuclear Information System (INIS)

    Lee, C. Y.; Jeong, J. M.; Kim, J. H.

    2010-01-01

    The mechanical property of the BaSrFBr:Eu phosphor layer of X-ray image plates was investigated by using image quality (IQ), resolution (LP/mm), and coincidence Doppler broadening (CDB) positron annihilation. The screen samples of BaSrFBr:Eu phosphors were irradiated with hospital X-rays in the course of diagnostic radiography at an average rate of 20,000 times per year and were used for various periods of time. The LP/mm values of the irradiated BaSrFBr:Eu image plates varied between 2.4 and 2.0 for three years while the IQ values varied between 35 and 11 over the same period. CDB positron annihilation spectroscopy was used to analyze the defect structures in the phosphor layer. The S parameter values increased in correlation with increased exposure time, which indicated that more defects were generated. There was a positive relationship between the IQ and S parameters. Measurements of the defects indicate that most of the defects were likely to have been generated by the X-ray radiation.

  11. Preliminary Validation of the MATRA-LMR Code Using Existing Sodium-Cooled Experimental Data

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Kim, Sangji

    2014-01-01

    The main objective of the SFR prototype plant is to verify TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. The fuel design limit is highly dependent on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, an accurate temperature calculation in each subassembly is highly important to assure a safe and reliable operation of the reactor systems. The current core thermalhydraulic design is mainly performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code, which has been already validated using the existing sodium-cooled experimental data. In addition to the SLTHEN code, a detailed analysis is performed using the MATRA-LMR (Multichannel Analyzer for Transient and steady-state in Rod Array-Liquid Metal Reactor) code. In this work, the MATRA-LMR code is validated for a single subassembly evaluation using the previous experimental data. The MATRA-LMR code has been validated using existing sodium-cooled experimental data. The results demonstrate that the design code appropriately predicts the temperature distributions compared with the experimental values. Major differences are observed in the experiments with the large pin number due to the radial-wise mixing difference

  12. Effects of applying three-dimensional seismic isolation system on the seismic design of FBR

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Yabana, Shuichi; Kanazawa, Kenji; Matsuda, Akihiro

    1997-01-01

    In this study conceptional three-dimensional seismic isolation system for fast breeder reactor (FBR) is proposed. Effects of applying three-dimensional seismic isolation system on the seismic design for the FBR equipment are evaluated quantitatively. From the evaluation, it is concluded following effects are expected by applying the three-dimensional seismic isolation system to the FBR and the effects are evaluated quantitatively. (1) Reduction of membrane thickness of the reactor vessel (2) Suppression of uplift of fuels by reducing vertical seismic response of the core (3) Reduction of the supports for the piping system (4) Three-dimensional base isolation system for the whole reactor building is advantageous to the combined isolation system of horizontal base isolation for the reactor building and vertical isolation for the equipment. (author)

  13. Sodium aerosol recovering device

    International Nuclear Information System (INIS)

    Fujimori, Koji; Ueda, Mitsuo; Tanaka, Kazuhisa.

    1997-01-01

    A main body of a recovering device is disposed in a sodium cooled reactor or a sodium cooled test device. Air containing sodium aerosol is sucked into the main body of the recovering device by a recycling fan and introduced to a multi-staged metal mesh filter portion. The air about against each of the metal mesh filters, and the sodium aerosol in the air is collected. The air having a reduced sodium aerosol concentration circulates passing through a recycling fan and pipelines to form a circulation air streams. Sodium aerosol deposited on each of the metal mesh filters is scraped off periodically by a scraper driving device to prevent clogging of each of the metal filters. (I.N.)

  14. Study of guided wave transmission through complex junction in sodium cooled reactor

    International Nuclear Information System (INIS)

    Elie, Q.; Le Bourdais, F.; Jezzine, K.; Baronian, V.

    2015-01-01

    Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presented in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)

  15. An evaluation of the fluid-elastic instability for Intermediate Heat Exchanger of Prototype Sodium-cooled fast Reactor

    International Nuclear Information System (INIS)

    Cho, Jaehun; Kim, Sungkyun; Koo, Gyeonghoi

    2014-01-01

    The sodium-cooled fast reactor (SFR) module consists of the vessel, containment vessel, head, rotating plug (RP), upper internal structure (UIS), intermediate heat exchanger (IHX), decay heat exchanger (DHX), primary pump, internal structure, internal components and reactor core. The IHXs transfer heat from the radioactive sodium coolant (primary sodium) in the primary heat transport system to the nonradioactive sodium coolant (secondary sodium) in the intermediate heat transport system. Each sodium flows like Fig. 1. Primary sodium flows inside of tube and secondary sodium flows outside. During transferring heat two sodium to sodium, the fluid-elastic instability is occurred among tube bundle by cross flow. Large amplitude vibration occurred by the fluid-elastic instability is caused such as crack and wear of tube. Thus it is important to decrease the fluid-elastic instability in terms of a safety. The purpose of this paper is to evaluate the fluid-elastic instability for tube bundle in the IHX following ASME code. This paper evaluated the fluid-elastic instability of tube bundle in the SFR IHX. According evaluation results, the fluid-elastic instability of IHX tube bundle is occurred. A installing an additional TSP under the upper tubesheet can decrease a probability of fluid-elastic instability. If a location of an additional TSP does not exceed tube length to become a 750 mm, tube bundle of IHX is safety from the fluid-elastic instability

  16. Seismic design for Monju FBR power plant

    International Nuclear Information System (INIS)

    1982-01-01

    This technical report introduces the basic concept on the aseismatic design of the FBR ''Monju'' power station, of which the construction in Tsuruga is planned by the Power Reactor and Nuclear Fuel Development Corp. The safety design of Monju has been performed according to ''The concept of evaluating the safety of fast breeder reactors'', and the thought concerning the aseismatic design also is written in it. According to it, ''The guide for the examination of aseismatic design regarding power reactor facilities'' should be referred to, and the classification according to the importance in aseismatic design must be made, taking the features in the design of liquid metal-cooled FBRs fully in consideration. In the aseismatic design of Monju performed according to these basic concept, the following two points were examined. In the aseismatic design of the equipment and piping, the difference of construction from LWRs such as low pressure, thin walled and high temperature construction is taken in consideration. The classification according to the aseismatic importance of the system and equipment is made on the basis of the features in the design of Monju. The classification according to aseismatic importance, the method of calculating earthquake power, the combination of loads and the allowable limit, and the aseismatic construction of the main facilities are reported. (Kako, I.)

  17. Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1975-01-01

    The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

  18. Methodology for Extraction of Remaining Sodium of Used Sodium Containers

    International Nuclear Information System (INIS)

    Jung, Minhwan; Kim, Jongman; Cho, Youngil; Jeong, Jiyoung

    2014-01-01

    Sodium used as a coolant in the SFR (Sodium-cooled Fast Reactor) reacts easily with most elements due to its high reactivity. If sodium at high temperature leaks outside of a system boundary and makes contact with oxygen, it starts to burn and toxic aerosols are produced. In addition, it generates flammable hydrogen gas through a reaction with water. Hydrogen gas can be explosive within the range of 4.75 vol%. Therefore, the sodium should be handled carefully in accordance with standard procedures even though there is a small amount of target sodium remainings inside the containers and drums used for experiment. After the experiment, all sodium experimental apparatuses should be dismantled carefully through a series of draining, residual sodium extraction, and cleaning if they are no longer reused. In this work, a system for the extraction of the remaining sodium of used sodium drums has been developed and an operation procedure for the system has been established. In this work, a methodology for the extraction of remaining sodium out of the used sodium container has been developed as one of the sodium facility maintenance works. The sodium extraction system for remaining sodium of the used drums was designed and tested successfully. This work will contribute to an establishment of sodium handling technology for PGSFR. (Prototype Gen-IV Sodium-cooled Fast Reactor)

  19. Sodium vapour aerosol formation and sodium deposition current work within the United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Hawtin, P [Chemical Engineering Division, Atomic Energy Research Establishment, Harwell, Didcot, Oxon (United Kingdom); Seed, G [Nuclear Power Company (Risley) Ltd, Risley, Warrington, Cheshire (United Kingdom)

    1977-01-01

    The significance to reactor operation of sodium transport through the cover gas of a sodium-cooled fast reactor and its subsequent deposition on cooled reactor surfaces is fully appreciated in the UK. A programme of work is therefore underway designed to understand the mechanism of sodium transport under these conditions. This paper described the work which has so far been completed, discussed the work presently in progress, and outlines future plans. (author)

  20. Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

    International Nuclear Information System (INIS)

    Wakai, Takashi; Machida, Hideo; Yoshida, Shinji; Xu, Yang; Tsukimori, Kazuyuki

    2014-01-01

    This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J IC , and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins

  1. Development of electro-magnetic pump for the ASTRID Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Suzuki, Tetsu; Aizawa, Rie; Wakasaki, Shingo; Dechelette, Frank; Benoit, Fabrice

    2017-01-01

    In the framework of the SFR (Sodium-cooled Fast Reactor) prototype called ASTRID (Advance Sodium Technological Reactor for Industrial Demonstration), the large capacity Electro-Magnetic Pumps (EMP) as main circulating pumps on the intermediate sodium circuits has been considered instead of mechanical pumps by CEA. The use of EMP has several decisive technological merits compared with mechanical pump in the reactor design, operation and maintenance. Nevertheless, some theoretical and technological developments have to be carried out in order to validate the design tools which take Magneto Hydro Dynamic (MHD) phenomena into account and the applicability of the EMP to the steady state and transient operating conditions of ASTRID. To move forward to developments, a collaboration agreement between the CEA and TOSHIBA Corporation was made and entered into to carry out a joint work program on the EMP for ASTRID design and development. CEA performed the theoretical analysis, and the EMP experimental model is constructed by CEA to support these theoretical developments. This model consists of a middle-size annular EMP for the liquid metal sodium. The various testing program using this model has been started in 2016. TOSHIBA performed the examination of design specification for ASTRID, an electromagnetic design, a structural design and various analyses. The structure design has been examined the placement of the sodium boundary and the withstand pressure, etc. And, if the thicknesses of the structure increase for withstanding pressure, the pump efficiency falls because the loss of the electromagnetic force increases. Therefore the balance between withstanding pressure and the efficiency has been considered by an electromagnetism design. This paper presents the design studies and experimental activities for the EMP development in the framework of the CEA-TOSHIBA collaborations. (author)

  2. Ultrasonic sensor for sodium perspective device

    International Nuclear Information System (INIS)

    Ogawa, Fujio; Onuki, Koji.

    1995-01-01

    The present invention concerns an ultrasonic wave sensor for a sodium perspective device disposed in an FBR type reactor, which can change the directing angle of the ultrasonic sensor irrespective of the external conditions in liquid sodium. Namely, the sensor comprises (1) a sensor main body, (2) a diaphragm disposed on an oscillating surface of ultrasonic waves generated from the sensor main body, (3) a pressurizing and depressurizing nozzle connected to the sensor main body, and (4) a pressure detector disposed to these nozzles. A gas is charged/discharged to and from the sensor main body to control a gas pressure in the main body. If the gas pressure is made higher, the diaphragm is deformed convexly. If the gas pressure is lowered, the diaphragm is deformed concavely. The directing angle is greater when it is deformed a convexly, and it is smaller when it is deformed concavely. Accordingly, ultrasonic wave receiving/sending range in the sodium can be varied optionally by controlling the gas pressure in the main body. (I.S.)

  3. JOYO modification program for demonstration tests of FBR innovative technology development

    International Nuclear Information System (INIS)

    Yoshimi, H.; Hachiya, Y.

    1990-01-01

    A plan is under way at PNC to modify the experimental fast reactor JOYO. The project is called MARK-III (MK-III) program. The purpose of MK-III is to expand the function of JOYO, and to make it possible to receive demonstration tests of new or high level technologies for FBR development. The MK-III program consists of two main modifications: conversion to a highly efficient irradiation facility; and a modification for demonstration testing of new technologies and concepts that have a high potential to reduce FBR plant construction cost, to evaluate plant reliability and to improve plant safety. These modifications are scheduled to start in 1991

  4. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

    2001-01-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  5. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

  6. Pilot and pilot-commercial plants for reprocessing spent fuels of FBR type reactors

    International Nuclear Information System (INIS)

    Shaldaev, V.S.; Sokolova, I.D.

    1988-01-01

    A review of modern state of investigations on the FBR mixed oxide uranium-plutonium fuel reprocessing abroad is given. Great Britain and France occupy the leading place in this field, operating pilot plants of 5 tons a year capacity. Technology of spent fuel reprocessing and specific features of certain stages of the technological process are considered. Projects of pilot and pilot-commercial plants of Great Britain, France, Japan, USA are described. Economic problems of the FBR fuel reprocessing are touched upon

  7. The transuranic mass balance during the introduction of metal fuel FBR cycle

    International Nuclear Information System (INIS)

    Yokoo, Takeshi; Inoue, Tadashi

    1999-01-01

    The mass flow of plutonium and minor actinides is calculated for a future light water reactor-fast breeder reactor (LWR-FBR) transition scenario, in which power generation by LWRs is continued on a certain scale for a long period before the replacement by FBRs begins. The burnup of the LWR spent fuel is considered to be higher than the current standard. It is assumed that all the plutonium and minor actinides recovered from LWRs are used to start up and feed metal fuel commercial FBRs, which replace those LWRs that have reached the end of their life. The results show that the accumulated plutonium and minor actinides from the LWRs can be consistently consumed without further accumulation, by gradually establishing the FBR power generation and its fuel cycle on the same scale. The optimum content of the minor actinides in the standard FBR fuel is about 2 weight percents. This result indicates that if FBRs are introduced in the future, extension of the LWR usage period will cause no significant problems in terms of the consumption of accumulated transuranic elements. (author)

  8. The current status of research and development concerning steam generator acoustic leak detection for the demonstration FBR plant

    International Nuclear Information System (INIS)

    Higuchi, Masahisa

    1990-01-01

    The Japan Atomic Power Co. (JAPC) started the research and development into Acoustic Leak Detection for the Demonstration FBR (D-FBR) plant in 1989. Acoustic Leak Detection is expected as a water leak detection system in the Steam Generator for the first D-FBR plant. JAPC is presently analyzing data on Acoustic Leak Detection in order to form some basic concepts and basic specifications about leak detection. Both low frequency types and high frequency types are selected as candidates for Acoustic Leak Detection. After a review of both types, either one will be selected for the D-FBT plant. A detailed Research and Development plan on Acoustic Leak Detection, which should be carried out prior to starting the construction of the D-FBR plant, is under review. (author). 3 figs, 2 tabs

  9. Assessment of the dry process fuel sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

  10. Assessment of the dry process fuel sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

  11. JOYO coolant sodium and cover gas purity control database (MK-II core)

    International Nuclear Information System (INIS)

    Ito, Kazuhiro; Nemoto, Masaaki

    2000-03-01

    The experimental fast reactor 'JOYO' served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon dioxide, methane and helium in argon gas with the reactor condition. (author)

  12. Basic nuclear data for FBR fuel cycle. Balance and forecasting

    International Nuclear Information System (INIS)

    Costa, L.; Granget, G.; Josso, F.

    1982-01-01

    A balance is made of nuclear data needed for studying FBR fuel cycle. From the accuracy of the obtained data, sensitivity calculations have enabled the future experimental measurements to be established [fr

  13. Status of sodium cooled fast reactors with closed fuel cycle in India

    International Nuclear Information System (INIS)

    Raj, B.

    2007-01-01

    Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

  14. Joint Feasibility study on the Instruction of a commercial FBR power plant to Korea between Korea and France

    International Nuclear Information System (INIS)

    Lee, Chang Kun; Cho, Mann; Juhn, Poong Eil; Nam, Jang Soo; Roh, Eun Rae; Yang, Chang Kook; Choi, Young Sang; Shin, Jae In; Chang, Young Nam; Kim, Hoa Gy; Lee, Hong Moon

    1985-03-01

    The objective of this joint feasibility study is to prepare for the implementation plan of the establishment of technological capability prior to the introduction of a commercial FBR power plant to Korea. The scope and contents of the second year project are as follows: Determined are the implementation activities in the two alternative FBR introduction scenarios i.e., a turnkey scenario and a gradual localization scenario with regard to the proportion of domestic participation. And FBR center is modeled as the construction of 4 units of 1,500 MWe LMFBR and related fuel cycle facilities such as reprocessing and fabrication plants at one site. In turnkey scenario, domestic implementation project is selected and the means to develop technical capability are classified into OJT/OJP, R and D, and technology transfer and the floating time schedule is made with reference to the first FBR operation year. In a gradual localization scenario Korean party's plan of means to develop technical capability is prepared. For the analysis of turnkey base introduction scenario project activities were extracted and analyzed referring to IAEA TR. 200 ''Manpower Development for Nuclear power Plant'', Guidebook (1980). And floating time schedule was prepared for the means to develop technical capability and project schedule. For the analysis of gradual localization scenario, project activities were derived under the assumption that entire facilities of FBR center should be localized. PWR's and FBR's fuel reprocessing facilities were incorporated in the reprocessing facility. (Author)

  15. Design study of an IHX support structure for a POOL-TYPE Sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

    2009-01-01

    The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity

  16. Utilization of cross-section covariance data in FBR core nuclear design and cross-section adjustment

    International Nuclear Information System (INIS)

    Ishikawa, Makoto

    1994-01-01

    In the core design of large fast breeder reactors (FBRs), it is essentially important to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of the plant. The cross-section errors, that is, covariance data are one of the most dominant sources for the prediction uncertainty of the core parameters, therefore, quantitative evaluation of covariance data is indispensable for FBR core design. The first objective of the present paper is to introduce how the cross-section covariance data are utilized in the FBR core nuclear design works. The second is to delineate the cross-section adjustment study and its application to an FBR design, because this improved design method markedly enhances the needs and importance of the cross-section covariance data. (author)

  17. Reliability of double-wall-tube steam generator for FBR considering water leak accident frequency

    International Nuclear Information System (INIS)

    Ueda, Nobuyuki; Kinoshita, Izumi; Nishi, Yoshihisa

    2000-01-01

    For early realization, a fast breeder reactor (FBR) is required to reduce construction cost. A reactor concept in which the intermediate heat transport system is eliminated by introducing a double-wall-tube steam generator is one convincing approach. The reliability of the double-wall-tube SG in a water leak accident (sodium-water reaction accident) due to tube failure is strongly related to the mitigating system design. The safety design of the double-wall-tube SG approach is investigated to limit the accident occurrence below 10 -7 (1/ry. A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to effectively prevent adhesion of the double-wall-tube. The reliability of the tube-to-tube plate was evaluated at 10 -10 (l/hr) for an inner tube and 10 -9 (l/hr) for an outer with reference to the failure experience of previous SGs. The failure must be detected within 30 to 60 minutes. (author)

  18. The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Philpott, E.F.; Pounder, F.; Willby, C.R.

    1978-01-01

    The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

  19. Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

    International Nuclear Information System (INIS)

    Chvetsov, I.; Kouznetsov, I.; Volkov, A.

    2000-01-01

    GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

  20. On the use of a moderation layer to improve the safety behavior in sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno, E-mail: b.merk@fzd.de [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany); Fridman, Emil; Weiss, Frank-Peter [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany)

    2011-05-15

    Research highlights: > Using a moderation layer can reduce the sodium void effect in a SFR. > Inserting the moderation layer improves the Doppler effect significantly. > The uniform layer distribution avoids effects on power and burnup distribution. > Hydride containing material like uranium-zirconium hydride is most efficient. - Abstract: This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B{sub 4}C or uranium-zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.

  1. FBR pellet fabrication - density and dimensional control

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1982-01-01

    The fuel pellet fabricating experience described in this paper involved pellet processing tests using mixed oxide (PuO 2 -UO 2 ) powders to produce fast breeder reactor (FBR) fuel pellets. Objectives of the pellet processing tests were to establish processing parameters for sintered-to-size fuel pellets to be used in an irradiation test in the Fast Flux Test Facility and to establish baseline fabrication control information. 26 figures, 7 tables

  2. Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Saxena, Aakanksha

    2014-01-01

    The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

  3. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  4. General purpose nonlinear analysis program FINAS for elevated temperature design of FBR components

    International Nuclear Information System (INIS)

    Iwata, K.; Atsumo, H.; Kano, T.; Takeda, H.

    1982-01-01

    This paper presents currently available capabilities of a general purpose finite element nonlinear analysis program FINAS (FBR Inelastic Structural Analysis System) which has been developed at Power Reactor and Nuclear Fuel Development Corporation (PNC) since 1976 to support structural design of fast breeder reactor (FBR) components in Japan. This program is capable of treating inelastic responses of arbitrary complex structures subjected to static and dynamic load histories. Various types of finite element covering rods, beams, pipes, axisymmetric, two and three dimensional solids, plates and shells, are implemented in the program. The thermal elastic-plastic creep analysis is possible for each element type, with primary emphasis on the application to FBR components subjected to sustained or cyclic loads at elevated temperature. The program permits large deformation, buckling, fracture mechanics, and dynamic analyses for some of the element types and provides a number of options for automatic mesh generation and computer graphics. Some examples including elevated temperature effects are shown to demonstrate the accuracy and the efficiency of the program

  5. Study of photo-stimulated luminescence in Ba0.95M0.05FBr:Eu2+ (M = Sr, Ca) powder

    International Nuclear Information System (INIS)

    Zhou Yingxue; Wang Dongsheng; Zhang Xinyi

    2001-01-01

    The luminescence centered at about 390 nm due to 4f 6 5d→4f 7 transition of Eu 2+ can be observed without X-ray or UV-light pre-irradiation for BaFBr:Eu 2+ , Ba 0.95 Sr 0.05 FBr:Eu 2+ and Ba 0.95 Ca 0.05 FBr:Eu 2+ powder samples which can be excited by light with wave length longer than 400 nm. It could be attributed to F centers which are formed in the preparation of samples. There are two broad bands in the photo-stimulated spectra. One peaked at 535 nm results from the excitation of electrons at F(F - ) center, and the other at 708 nm might be originated from the electron excitation of F(Br - ) center or F(Br - ) accumulators. If the Ba 2+ (ion's radius equals to 0.135 nm) in the samples were replaced by about 50% mol. of Sr 2+ (ion's radius equals to 0.113 nm) or Ca 2+ (ion's radius equals to 0.099 nm), the intensities of 535 nm and 708 nm bands decrease with the decreasing of ion radius. On the other hand, the shoulders at 580 nm and 650 nm or 575 nm and 645 nm can appear. In absorption spectra of 50-400 nm range, one can observe the red-shift of 79 nm, 83 nm peaks and new absorption peaks. It means that the new color centers: F(F - , Sr 2+ ), F(br - , Sr 2+ ) or F(F - , Ca 2+ ), F(br - , Ca 2+ ) are formed, when Ba 2+ were replaced with Sr 2+ or Ca 2+ in samples. It corresponds to the shoulders in photo-stimulated spectra

  6. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  7. IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

  8. Proposal of a nuclear cycle research and development plan in Tokai works. The roadmap from LWR cycle to FBR cycle

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Abe, Tomoyuki; Kashimura, Takuo; Nagai, Toshihisa; Maeda, Seichiro; Yamaguchi, Toshiya; Kuroki, Ryoichiro

    2003-07-01

    The Generation-II Project Task Force Team has investigated a research and development plan of a future nuclear fuel cycle in Tokai works for about three months from December 19, 2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R and D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. Establishment and advancement of 'the tough nuclear fuel cycle'. Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste. And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology. (author)

  9. Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

  10. Small leak detection by measuring surface oscillation during sodium-water reaction in steam generator

    International Nuclear Information System (INIS)

    Nei, Hiromichi; Hori, Masao

    1977-01-01

    Small leak sodium-water reaction tests were conducted to develop various kinds of leak detectors for the sodium-heated steam generator in FBR. The super-heated steam was injected into sodium in a reaction vessel having a sodium free surface, simulating the steam generator. The level gauge in the reaction vessel generated the most reliable signal among detectors, as long as the leak rates were relatively high. The level gauge signal was estimated to be the sodium surface oscillation caused by hydrogen bubbles produced in sodium-water reaction. Experimental correlation was derived, predicting the amplitude as a function of leak rate, hydrogen dissolution ratio, bubble rise velocity and other parameters concerned, assuming that the surface oscillation is in proportion to the gas hold-up. The noise amplitude under normal operation without water leak was increased with sodium flow rate and found to be well correlated with Froud number. These two correlations predict that a water leak in a ''MONJU'' class (300 MWe) steam generator could possibly be detected by level gauges at a leak rate above 2 g/sec. (auth.)

  11. Under sodium ultrasonic viewing for Fast Breeder Reactors: a review

    International Nuclear Information System (INIS)

    Tarpara, Eaglekumar G.; Patankar, V.H.; Vijayan Varier, N.

    2016-09-01

    Liquid Metal Fast Breeder Reactors (LMFBR/FBR) are of two types: Loop type and Pool type. Many countries like USA, Japan, UK, Russia, China, France, Lithuania, Belgium, Korea, and India have worked extensively on these types of FBRs. FBRs are capable of breeding more fissionable fuel than it consumes like breeding of Plutonium-239 from non-fissionable Uranium-238. In FBR, heat is released by fission process and it must be captured and transferred to the electric generator by the liquid metal coolant (i.e. Sodium). Due to continuous operation and for safety and licensing reasons, periodic inspection and maintenance is required for reactor fuel assemblies which carry nuclear fuels. For this reason, under sodium ultrasonic imaging technique is adopted as in-service inspection activity for viewing of core of FBRs. Since liquid sodium is optically opaque, ultrasonic technique is the only method which can be employed for imaging in liquid sodium. In harsh conditions like high temperature and high radiation, there is a restriction on the development of possible ultrasonic visualization systems and selection of the transducer materials which can operate in the core region of reactor at around 200 °C during shutdown of reactor. This report provides a review of works related to ultrasonic imaging in sodium, different materials used in high temperature transducer assemblies and their different coupling/bonding techniques to achieve maximum transmission efficiency in high temperature sodium environment. The report also provides the overview of different architectures and imaging methods of transducer array elements which were used in LMFBRs for inspection and visualization of the reactor core sub-assemblies. The report is focused on a review of some possible beam forming techniques which may be used for nuclear applications for high temperature environment. Published information of the different simulation models are also reviewed which can be adopted to simulate the

  12. BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Hetrick, D.L.; Sowers, G.W.

    1978-06-01

    This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices

  13. Characterization of a sodium-cooled fast reactor in an MHR-SFR synergy for TRU transmutation

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Kim, Yonghee; Venneri, Francesco

    2008-01-01

    In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50-65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR-SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core

  14. Conceptual design of advanced central receiver power systems sodium-cooled receiver concept. Volume 2, Book 2. Appendices. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-01

    The appendices include: (A) design data sheets and P and I drawing for 100-MWe commercial plant design, for all-sodium storage concept; (B) design data sheets and P and I drawing for 100-MWe commercial plant design, for air-rock bed storage concept; (C) electric power generating water-steam system P and I drawing and equipment list, 100-MWe commercial plant design; (D) design data sheets and P and I drawing for 281-MWe commercial plant design; (E) steam generator system conceptual design; (F) heat losses from solar receiver surface; (G) heat transfer and pressure drop for rock bed thermal storage; (H) a comparison of alternative ways of recovering the hydraulic head from the advanced solar receiver tower; (I) central receiver tower study; (J) a comparison of mechanical and electromagnetic sodium pumps; (K) pipe routing study of sodium downcomer; and (L) sodium-cooled advanced central receiver system simulation model. (WHK)

  15. Study of various Brayton cycle designs for small modular sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Lee, Jeong Ik

    2014-01-01

    Highlights: • Application of closed Brayton cycle for small and medium sized SFRs is reviewed. • S-CO 2 , helium and nitrogen cycle designs for small modular SFR applications are analyzed and compared in terms of cycle efficiency, component performance and physical size. • Several new layouts for each Brayton cycle are suggested to simplify the turbomachinery designs. • S-CO 2 cycle design shows the best efficiency and compact size compared to other Brayton cycles. - Abstract: Many previous sodium cooled fast reactors (SFRs) adopted steam Rankine cycle as the power conversion system. However, the concern of sodium water reaction has been one of the major design issues of a SFR system. As an alternative to the steam Rankine cycle, several closed Brayton cycles including supercritical CO 2 cycle, helium cycle and nitrogen cycle have been suggested recently. In this paper, these alternative gas Brayton cycles will be compared to each other in terms of cycle performance and physical size for small modular SFR application. Several new layouts are suggested for each fluid while considering the turbomachinery design and the total system volume

  16. Calculation of the neutron noise induced by periodic deformations of a large sodium-cooled fast reactor core

    International Nuclear Information System (INIS)

    Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.

    2014-01-01

    The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)

  17. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  18. Review of aerosol problems and the theory of aerosol physics with particular reference to sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Williams, R.J.

    1978-01-01

    Processes that would govern the development, transport, and removal of aerosols, which are of interest in the study of hypothetical core disruptive situations in pool type sodium cooled fast reactors, are discussed. Theoretical descriptions of these processes are presented and known inadequacies indicated. The interpretation of experimental data and numeric solution of the governing equations is briefly considered. (author)

  19. Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwai, Takehiko

    1998-07-01

    A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

  20. Present condition of survey research on actualization strategy of fast breeding reactor (FBR) cycling. General outlines on the research

    International Nuclear Information System (INIS)

    Hara, Hideaki

    2001-01-01

    The Japan Nuclear Cycle Development Institute (JNC) started the survey research on actualization strategy of FBR cycling under cooperation of related organizations such as electric business company and so on, on July, 1999. The research aims at preparation of technical system to establish the FBR cycling for a future main energy supply source by extracting an actualization picture maximum activated advantages originally haven by the FBR cycling and by proposing a developmental strategy flexibly responsible to diverse needs in future society. Here was reported on effort state of its phase 1 (two years between 1999 and 2000 fiscal years). In the phase 1, it was planned to perform research and development shown as follows: 1) Extraction of actualization candidate concept on the FBR cycling under a premise of safety security and a viewpoint of evaluation on economics, resource effective usage, environmental loading reduction, and nuclear dispersion resistance by conducting investigation and evaluation of wide technical choices adopting innovative techniques, and 2) Embodiment of a research and development program of phase 2 (from 2001 to 2005 fiscal years) by investigating some technical subjects important for selection of research and development program aiming at actualization and its candidate concept on the FBR cycling. (G.K.)

  1. Development of knowledge-based operator support system for steam generator water leak events in FBR plants

    International Nuclear Information System (INIS)

    Arikawa, Hiroshi; Ida, Toshio; Matsumoto, Hiroyuki; Kishida, Masako

    1991-01-01

    A knowledge engineering approach to operation support system would be useful in maintaining safe and steady operation in nuclear plants. This paper describes a knowledge-based operation support system which assists the operators during steam generator water leak events in FBR plants. We have developed a real-time expert system. The expert system adopts hierarchical knowledge representation corresponding to the 'plant abnormality model'. A technique of signal validation which uses knowledge of symptom propagation are applied to diagnosis. In order to verify the knowledge base concerning steam generator water leak events in FBR plants, a simulator is linked to the expert system. It is revealed that diagnosis based on 'plant abnormality model' and signal validation using knowledge of symptom propagation could work successfully. Also, it is suggested that the expert system could be useful in supporting FBR plants operations. (author)

  2. Experimental thermal hydraulics in support of FBR

    International Nuclear Information System (INIS)

    Padmakumar, G.; Anand Babu, C.; Kalyanasundaram, P.; Vaidyanathan, G.

    2009-01-01

    The thermal hydraulic design plays a crucial role for the safe and economical deployment of Liquid Metal Cooled Fast Breeder Reactor (LMFBR). Robust experimental programmes are required in support of LMFBR thermal hydraulics design. The philosophy of testing has been to construct small scale models to understand the physical behaviour and to build larger scale models to optimize the component design. The experiments are conducted either in sodium or using a simulant like water/air. The paper gives a brief account of the various thermal hydraulic experiments carried out in support of the design of Prototype Fast Breeder Reactor (PFBR). (author)

  3. Ultrasonic sweep arm for sodium cooled reactors

    International Nuclear Information System (INIS)

    Rohrbacher, H.A.; Bartholomay, R.

    1975-05-01

    This report describes experience in the use of a new type of monitoring and testing device to be applied in conjunction with components under sodium. In the method outlined, ultrasonic pulses are used which are emitted into the sodium plenum of fast breeder reactors by newly developed high temperature transducers. The basic work was conducted under out-of-pile conditions in a sodium tank of the sodium tank facility of the Karlsruhe Institute for Reactor Development. The sensor development, which preceded this phase, resulted in the use of soldered lithium niobate crystals whose operating characteristics were improved by the preliminary treatment outlined in the report. Special materials and techniques suitable for sensor fabrication are proposed. An alternative to soldering is suggested for contacting the crystals with their diaphragms, i.e. a contact pressure concept for the range of application up to 2 MHz. (orig.) [de

  4. Earthquake-proof support structures for the recycling pump in FBR type reactors

    International Nuclear Information System (INIS)

    Nakagawa, Masaki; Shigeta, Masayuki.

    1984-01-01

    Purpose: To improve the earthquake proofness of the recycling pump for use in FBR type reactors upon earthquake by reducing the vibration response of the pump. Constitution: The outer casing of a recycle pump suspended into liquid sodium is extended to the portion that penetrates a reactor core support structures. Support structures surrounding the outer side of the recycling pump are disposed with a gap not restraining the free thermal deformations of the recycling pump to the inside of the partition wall structures and the portion of the recycling pump penetrating the reator core support structures, to integrate the support structures with the reactor core support structures. Accordingly, there are no interferences between the recycling pump and the support structures with respect to the thermal deformations that change gradually with time. Upon vibrating under the rapidly changing external forces of earthquakes, however, the pressure resulted to the liquid in the gap due to the vibrations of the recycling pump is transmitted with no escape to the support structures, the recycling pump and the support structures integrally resist the vibrations thereby enabling to reduce the vibrations in the recycling pumps. (Horiuchi, T.)

  5. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  6. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  7. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  8. Study on flow-induced vibration of large-diameter pipings in a sodium-cooled fast reactor. Influence of elbow curvature on velocity fluctuation field

    International Nuclear Information System (INIS)

    Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira

    2010-02-01

    The main cooling system of Japan Sodium-cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9m/s in the primary hot leg pipe with large-diameter (1.3m). The maximum Reynolds number in the piping reaches 4.2x10 7 . The hot leg pipe having a 90 degree elbow with curvature ratio of r/D=1.0, so-called 'short elbow', which enables a compact reactor vessel. In sodium cooled fast reactors, the system pressure is so low that thickness of pipings in the cooling system is thinner than that in LWRs. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned at the short elbow. The evaluation of the structural integrity of pipings in JSFR should be conducted based on a mechanistic approach of FIV at the elbow. It is significant to obtain the knowledge of the fluctuation intensity and spectra of velocity and pressure fluctuations in order to grasp the mechanism of the FIV. In this study, water experiments were conducted. Two types of 1/8 scaled elbows with different curvature ratio, r/D=1.0, 1.5, were used to investigate the influence of curvature on velocity fluctuation at the elbow. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow at the elbow outlet and separation flow at the inner wall of elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow. (author)

  9. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  10. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  11. Effect of moving distance of temperature distribution on thermal ratchetting behavior of a FBR reactor vessel

    International Nuclear Information System (INIS)

    Ueta, Masahiro; Douzaki, Kouji; Takahashi, Yukio; Ooka, Yuji; Osaki, Toshio; Take, Kouji.

    1992-01-01

    It should be considered in a FBR reactor vessel design that thermal ratchetting might be caused by moving axial thermal gradient, in other words, moving sodium level. The behavior and the mechanism of ratchetting have almost become clear by studies for the past several years. A simplified evaluation method for ratchetting behavior has been proposed. However, the evaluation method has been shown to be excessively conservative by testing results. In this paper, the effect of moving distance of axial temperature distributions, which is one of main factors to be considered in precise estimation of ratchetting behavior, is studied by inelastic analyses. Based on the study, it is proposed to introduce a strain reducing factor taking account of residual stresses in the region of moving axial temperature distribution to the original evaluation method. The new method has been validated by comparing the prediction with results of both testing and the original method. (author)

  12. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  13. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  14. Development of radioactivity estimation system considering radioactive nuclide movement

    International Nuclear Information System (INIS)

    Fukumura, Nobuo; Miyamoto, Yoshiaki

    2010-01-01

    A radioactivity estimation system considering radioactive nuclide movement is developed to integrate the established codes and the code system for decommissioning of sodium cooled fast reactor (FBR). The former are the codes for estimation of radioactivity movement in sodium coolant of fast reactor which are named SAFFIRE, PSYCHE and TTT. The latter code system is to estimate neutron irradiation activity (COSMARD-RRADO). It is paid special attention to keep the consistency of input data used among these codes and also the simplification of their interface. A new function is added to the estimation system, to estimate minor FP inventory caused by the fission of impurities contained in the coolant and slight fuel material attached on the fuel cladding. To check the evaluation system, the system is applied with radioactivity data of the preceding FBR such as BN-350, JOYO and Monju. Agreement between the analysis results and the measurement is well satisfactory. The uncertainty of the code system is within several tens per cent for the activation of primary coolant (Na-22) and factor of 2-4 for the estimation of radioactivity inventory in sodium coolant. (author)

  15. Effect of meat ingredients (sodium nitrite and erythorbate) and processing (vacuum storage and packaging atmosphere) on germination and outgrowth of Clostridium perfringens spores in ham during abusive cooling.

    Science.gov (United States)

    Redondo-Solano, Mauricio; Valenzuela-Martinez, Carol; Cassada, David A; Snow, Daniel D; Juneja, Vijay K; Burson, Dennis E; Thippareddi, Harshavardhan

    2013-09-01

    The effect of nitrite and erythorbate on Clostridium perfringens spore germination and outgrowth in ham during abusive cooling (15 h) was evaluated. Ham was formulated with ground pork, NaNO2 (0, 50, 100, 150 or 200 ppm) and sodium erythorbate (0 or 547 ppm). Ten grams of meat (stored at 5 °C for 3 or 24 h after preparation) were transferred to a vacuum bag and inoculated with a three-strain C. perfringens spore cocktail to obtain an inoculum of ca. 2.5 log spores/g. The bags were vacuum-sealed, and the meat was heat treated (75 °C, 20 min) and cooled within 15 h from 54.4 to 7.2 °C. Residual nitrite was determined before and after heat treatment using ion chromatography with colorimetric detection. Cooling of ham (control) stored for 3 and 24 h, resulted in C. perfringens population increases of 1.46 and 4.20 log CFU/g, respectively. For samples that contained low NaNO2 concentrations and were stored for 3 h, C. perfringens populations of 5.22 and 2.83 log CFU/g were observed with or without sodium erythorbate, respectively. Residual nitrite was stable (p > 0.05) for both storage times. Meat processing ingredients (sodium nitrite and sodium erythorbate) and their concentrations, and storage time subsequent to preparation of meat (oxygen content) affect C. perfringens spore germination and outgrowth during abusive cooling of ham. Copyright © 2013 Elsevier Ltd. All rights reserved.

  16. Proposals for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1976-01-01

    Design and operational experience of CEGB gas cooled reactors and certain overseas reactor plant is reviewed in relation to in-service inspection and monitoring capabilities. Design guidelines and preliminary proposals are given for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors. Specific comments are made on the items of further design and development work believed to be necessary

  17. Level-1 PSA to support the design of the KALIMER-600 Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Han, Sang Hoon; Kim, Tae-Woon; Jeong, Hae-Yong; Han, Seok Joong; Ahn, Kwang-Il; Yang, Joon-Eon

    2012-01-01

    A sodium-cooled fast reactor, KALIMER-600, is under development. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as a coolant. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing KALIMER-600 from the aspects of risk informed design. A preliminary level-1 internal full power PSA has been performed to evaluate the safety level and its applicability for the KALIMER-600 conceptual design. Various design alternatives are evaluated from the viewpoint of PSA in order to support the design of the KALIMER-600. Sensitivity studies are also performed to evaluate the assumptions made for the PSA. The applicability and weakness of the KALIMER-600 PSA are discussed. The technical issues to be solved in performing the PSA will be discussed. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  19. Construction within cooling system of a sodium cooler reactor

    International Nuclear Information System (INIS)

    1977-01-01

    A procedure is described for the manufacture and the construction of a bundle of a large number of pipes, at least near their outer ends lying practically evenly spaced which pipes lie with one of their outermost ends in a pipe plate and with their other outer ends in a second pipe plate, where the procedure involves placing at or near the derived place a means for holding the bundle of pipes, as well as eventually holding a pipe plate with stub pipes near the outer ends of the bundle of pipes, the successive attachment by means of welding of the pipes in the plate of the above mentioned assembly with the stub pipes, characterized in that to each of the pipes in the bundle is welded to an outer end directly a corresponding short pipe which is also welded to a pipe end of a stub pipe, so that a connection is made by the short pipe which lies between the outer end of the pipe in the bundle and the stub pipe. Such a construction is used in the heat exchanger of sodium cooled reactors. (G.C.)

  20. Linear programming optimization of nuclear energy strategy with sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Lee, Je Whan; Jeong, Yong Hoon; Chang, Yoon Il; Chang, Soon Heung

    2011-01-01

    Nuclear power has become an essential part of electricity generation to meet the continuous growth of electricity demand. A Sodium-cooled Fast Reactor (SFR) was developed to extend uranium resource utilization under a growing nuclear energy scenario while concomitantly providing a nuclear waste management solution. Key questions in this scenario are when to introduce SFRs and how many reactors should be introduced. In this study, a methodology using Linear Programming is employed in order to quantify an optimized growth pattern of a nuclear energy system comprising light water reactors and SFRs. The optimization involves tradeoffs between SFR capital cost premiums and the total system U3O8 price premiums. Optimum nuclear growth patterns for several scenarios are presented, as well as sensitivity analyses of important input parameters

  1. Monitoring Sodium Circuits and ACSR cables using Fiber Optic Sensors

    International Nuclear Information System (INIS)

    Kasinathan, M.; Sosamma, S.; Babu-Rao, C.; Kumar, Anish; Purna-Chandra-Rao, B.; Murali, N; Jayakumar, T.

    2013-06-01

    Raman Distributed Temperature Sensors (RDTS) are attractive for the monitoring of coolant loop systems in nuclear power plants and monitoring of overhead power transmission lines. This paper discusses deployment of RDTS on double walled pipelines of primary sodium circuits in Fast Breeder Reactors (FBR). It is demonstrated as a proof-of-concept on a test loop with water as the leaking medium. Path delay multiplexing is adopted to improve the spatial resolution from 1.02 m to 0.5 m. A second application focuses on the influence of environmental factors on the detectability of defects in the ACSR cables using RDTS. (authors)

  2. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  3. Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

    Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

  4. Simulation of Fungal-Mediated Cell Death by Fumonisin B1 and Selection of Fumonisin B1–Resistant (fbr) Arabidopsis Mutants

    Science.gov (United States)

    Stone, Julie M.; Heard, Jacqueline E.; Asai, Tsuneaki; Ausubel, Frederick M.

    2000-01-01

    Fumonisin B1 (FB1), a programmed cell death–eliciting toxin produced by the necrotrophic fungal plant pathogen Fusarium moniliforme, was used to simulate pathogen infection in Arabidopsis. Plants infiltrated with 10 μM FB1 and seedlings transferred to agar media containing 1 μM FB1 develop lesions reminiscent of the hypersensitive response, including generation of reactive oxygen intermediates, deposition of phenolic compounds and callose, accumulation of phytoalexin, and expression of pathogenesis-related (PR) genes. Arabidopsis FB1-resistant (fbr) mutants were selected directly by sowing seeds on agar containing 1 μM FB1, on which wild-type seedlings fail to develop. Two mutants chosen for further analyses, fbr1 and fbr2, had altered PR gene expression in response to FB1. fbr1 and fbr2 do not exhibit differential resistance to the avirulent bacterial pathogen Pseudomonas syringae pv maculicola (ES4326) expressing the avirulence gene avrRpt2 but do display enhanced resistance to a virulent isogenic strain that lacks the avirulence gene. Our results demonstrate the utility of FB1 for high-throughput isolation of Arabidopsis defense-related mutants and suggest that pathogen-elicited programmed cell death of host cells may be an important feature of compatible plant–pathogen interactions. PMID:11041878

  5. PAPIRUS - a computer code for FBR fuel performance analysis

    International Nuclear Information System (INIS)

    Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

    1991-01-01

    The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

  6. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  7. Preliminary study on flexible core design of super FBR with multi-axial fuel shuffling

    International Nuclear Information System (INIS)

    Sukarman; Yamaji, Akifumi; Someya, Takayuki; Noda, Shogo

    2017-01-01

    Preliminary study has been conducted on developing a new flexible core design concept for the Supercritical water-cooled Fast Breeder Reactor (Super FBR) with multi-axial fuel shuffling. The proposed new concept focuses on the characteristic large axial coolant density change in supercritical water cooled reactors (SCWRs) when the coolant inlet temperature is below the pseudocritical point and large coolant enthalpy rise is taken in the core for achieving high thermal efficiency. The aim of the concept is to attain both the high breeding performance and good thermal-hydraulic performance at the same time. That is, short Compound System Doubling Time (CSDT) for high breeding, large coolant enthalpy rise for high thermal efficiency, and large core power. The proposed core concept consists of horizontal layers of mixed oxide (MOX) fuels and depleted uranium (DU) blanket layers at different elevation levels. Furthermore, the upper core and the lower core are separated and independent fuel shuffling schemes in these two core regions are considered. The number of fuel batches and fuel shuffling scheme of the upper core were changed to investigate influence of multi-axial fuel shuffling on the core characteristics. The core characteristics are evaluated with-three-dimensional diffusion calculations, which are fully-coupled with thermal-hydraulics calculations based on single channel analysis model. The results indicate that the proposed multi-axial fuel shuffling scheme does have a large influence on CSDT. Further investigations are necessary to develop the core concept. (author)

  8. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  9. Analysis of self-wastage phenomena of micro leak caused by sodium-water reaction in sodium-cooled fast breeder reactor through simulant experiment

    International Nuclear Information System (INIS)

    Jang, Sunghyon; Takata, Takashi; Yamaguchi, Akira

    2014-01-01

    Self-wastage phenomena are an enlargement of a leak on the heat transfer tube caused by a corrosive sodium-water reaction (SWR) in a steam generator (SG) of sodium-cooled fast breeder reactor (SFR). If the steam generator operates for sometimes under this condition, the self-wastage phenomena start from the sodium side and advance through the tube thickness. The leak rate stays almost constant level until the wastage reaches the sodium side, however, when the thin diaphragm of the tube wall is removed, the leak rate sharply increase, and it may bring a secondary failure of the surrounding heat transfer tubes. The design and safety concern is a possibility of the secondary failure of nearby SG tubes that could cause undesirable development of the accidents. One needs to evaluate the increased resultant leak rate due to the self-wastage phenomenon. Therefore, a quantification of the diameter of enlarged leak is needed to estimate the resultant leak rate. For this purpose, a simulant self-wastage experiment was proposed to investigate the self-enlargement of the leak so that evaluate the mechanism of the Self-wastage. In the experiment, high concentrated hydrochloric acid (HCl) is injected to the reaction tank that is filled sodium hydroxide (NaOH) solution through a nozzle made by paraffin wax. The self-enlargement of the leak was evaluated by considering the melted nozzle due to the reaction heat released from the Neutralization reaction. Also, a numerical investigation has been carried out to evaluate the enlarged nozzle and validate the results of experimental methodology. Based on the experimental and computational results, it is found that despite initial leak rate, there is an upper limit in the enlarged nozzle. These results show a similar tendency with the experimental result of SWAT-4 experiment carried out by Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan. Furthermore, the increased resultant leak rate is evaluated using the enlarged

  10. Numerical study on pressure drop and heat transfer for designing sodium-to-air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, Hie-Chan; Eoh, Jae-Hyuk; Cha, Jae-Eun; Kim, Seong-O.

    2013-01-01

    Highlights: ► Numerical simulation for the heat flow characteristic of the sodium-to-air heat exchanger (AHX) and tube banks. ► Parallelogram tube banks showed almost similar thermal and hydraulic characteristics to the rectangular tube banks. ► Pressure drop and heat transfer of the staggered and rectangular tube banks compared with Zhukauskas’ correlation. ► AHX was modeled as porous media and suggested design guide to enhance the performance. - Abstract: A numerical study is performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX are modeled as porous media and simulated heat and momentum transfer by a commercial program. Two-dimensional flow characteristic appears differently at the inlet region of the AHX annulus, and the required length of the inlet region is shorter for an inlet having a 45 degree chamber or a round shape than for one with a perpendicular corner. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX are evaluated and discussed. Pressure drop and heat transfer shows similar trends and underestimated values, respectively, when compared with Zhukauskas empirical correlations. The parallelogram tube bank shows similar results to the rectangular arrangement.

  11. Sodium fire suppression

    International Nuclear Information System (INIS)

    Malet, J.C.

    1979-01-01

    Ignition and combustion studies have provided valuable data and guidelines for sodium fire suppression research. The primary necessity is to isolate the oxidant from the fuel, rather than to attempt to cool the sodium below its ignition temperature. Work along these lines has led to the development of smothering tank systems and a dry extinguishing powder. Based on the results obtained, the implementation of these techniques is discussed with regard to sodium fire suppression in the Super-Phenix reactor. (author)

  12. Sodium fire suppression

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J C [DSN/SESTR, Centre de Cadarache, Saint-Paul-lez-Durance (France)

    1979-03-01

    Ignition and combustion studies have provided valuable data and guidelines for sodium fire suppression research. The primary necessity is to isolate the oxidant from the fuel, rather than to attempt to cool the sodium below its ignition temperature. Work along these lines has led to the development of smothering tank systems and a dry extinguishing powder. Based on the results obtained, the implementation of these techniques is discussed with regard to sodium fire suppression in the Super-Phenix reactor. (author)

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

    International Nuclear Information System (INIS)

    1984-06-01

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

  14. Reactor structure for FBR

    International Nuclear Information System (INIS)

    Yamazaki, Tsukasa.

    1986-01-01

    Purpose: To prevent deformation of equipment for FBR structure by inhibiting free convection generated at the roof slab through device. Constitution: The labyrinth is placed between the lower part of the roof slab and the lower one of the positioning flange, and then, convection-preventive wrinkle is provided for the side wall for the positioning flange against the roof slab side wall. The upper part of the positioning flange is fixed to the upper surface of the roof slab, the through-device flange is connected to the lower flange, and prevent generation of thermal stress. Thus, free convection at the through-device is prevented, and it has become possible to miniaturize the seal section of the intermediate heat exchanger and prevent galling of the circulating pump. The joint position of the positioning flange with the through-device flange can be shifted to the same height level of the roof slab, and the length under the hook of the overhead crance can be reduced. (Horiuchi, T.)

  15. Sodium cleaning device for nuclear reactor equipments

    International Nuclear Information System (INIS)

    Fujisawa, Morio.

    1985-01-01

    Purpose: To enable sodium cleaning over the entire length of large size equipments such as control rod drives and primary coolant recycling pumps for use in FBR type reactors. Constitution: A plurality of warm water supply nozzles each having a valve are connected at varying height on the side of a cleaning tank, to which an exhaust line is connected. These nozzles are connected with an exhaust port at the bottom of the tank to constitute a pipeway for cleaning warm water recycling line including a water feed pump and a feedwater heater. The water level in the tank is changed stepwise by successively selecting the warm water feed nozzles. Further, the warm water in the tank is recyclically fed through the nozzles selected at each step of the water level through the recycle line while warming. On the other hand, the pressure inside the tank is reduced through the exhaust line, whereby the warm water in the tank is boiled at low temperature to clean-up sodium on the equipments to be cleaned over the entire length. (Horiuchi, T.)

  16. Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

    1992-01-01

    In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

  17. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  18. Development of sodium leak detection technology using laser resonance ionization mass spectrometry. Design and functional test using prototype sodium detection system

    International Nuclear Information System (INIS)

    Aoyama, Takafumi; Ito, Chikara; Harano, Hideki; Okazaki, Koki; Watanabe, Kenichi; Iguchi, Tetsuo

    2009-01-01

    In a sodium-cooled fast reactor, highly sensitive technology is required to detect small amounts of sodium leaking from the cooling system piping or components. The conventional sodium leak detectors have a fundamental difficulty in improving the detection sensitivity for a sodium leak because of the presence of salinity ( 23 NaCl) in the atmosphere around the components and piping of cooling systems. In order to overcome this problem, an innovative technology has been developed to selectively detect the radioactive sodium ( 22 Na) produced by a neutron reaction in the primary cooling system using Laser Resonance Ionization Mass Spectrometry (RIMS). In this method, sodium ions produced with the two processes of (1) atomization of sodium aerosols and (2) resonance ionization of sodium atom, are detected selectively using a time-of-flight mass spectrometer. The 22 Na can be distinguished from the stable isotope ( 23 Na) by mass spectrometry, which is the advantage of RIMS comparing to the other methods. The design and the construction of the prototype system based on fundamental experiments are shown in the paper. The aerodynamic lens was newly introduced, which can transfer aerosols at atmospheric pressure into a vacuum chamber while increasing the aerosol density at the same time. Furthermore, the ionization process was applied by using the external electric field after resonance exciting from the ground level to the Rydberg level in order to increase the ionization efficiency. The preliminary test results using the stable isotope ( 23 Na) showed that prototype system could easily detect sodium aerosol of 100 ppb, equivalent to the sensitivity of the conventional detectors. (author)

  19. Neutron noise analysis for malfunction diagnosis at sodium cooled reactors

    International Nuclear Information System (INIS)

    Hoppe, P.

    1978-09-01

    For the investigation of the potential use of neutron noise analysis at sodium cooled power reactors, measurements have been performed at the KNK I reactor over a period of 18 month under different operational conditions. The signal fluctuations of the following tranducers have been recorded: In-core and Ex-core neutron detectors, temperature-, flow-, pressure-, vibration- and acoustic sensors. These extensive measurements have been analyzed in the frequency range from 0,001 Hz to 1000 Hz with all currently known methods for the identification of noise sources. The following results have been found: - Neutron noise for f 20 Hz the white detection noise prevails. In the region from 1 Hz to 20 Hz the vibrations of core components contribute to neutron noise. - Neutron noise is influenced by the state of the plant. - The contributions to neutron noise due to the fluctuations of coolant flow and inlet temperature are small compared to those produced by the movements of the control rod initiated by the reactor control system. The quantitatively unidentifiable amount of reactivity fluctuations (0,6 time-dependent thermal bowing of the core. With respect to these results and by calculation of the neutron noise patterns to be expected for the SNR 300, the following possible applications for neutron noise analysis have been found: By means of neutron noise analysis only reactivity fluctuations can be identified and supervised which are produced by time dependent changes of the core geometry. Furthermore neutron noise analysis is well suited for a sensitive detection of control rod vibrations and of local sodium boiling. Finally it can be used for the surveillance of the proper functioning of the reactor control system and of the control rod drive mechanism. (orig./HP) 891 HP [de

  20. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube

  1. Interrelationship betwen material strength and component design under elevated temperature for FBR

    International Nuclear Information System (INIS)

    Nakagawa, Y.

    Structural design under elevated temperature for fast breeder reactor plant is very troublesome compared to that of for lower temperature. This difficulty can be mainly discussed from two different stand points. One is design and design code, another is material strength. Components in FBR are operated under creep regime and time dependent creep behaviour should be elevated properly. This means the number and combinations of design code and material strength are significantly large and makes these systems very complicated. Material selection is, in no words, not an easy job. This should be done by not only material development but also component design stand point. With valuable experience of construction and research on FBR, a lot of information on component design and material behaviour is available. And it is a time to choose the ''best material'' from the entire stand points of component construction. (author)

  2. Replacement inhibitors for tank farm cooling coil systems

    International Nuclear Information System (INIS)

    Hsu, T.C.

    1995-01-01

    Sodium chromate has been an effective corrosion inhibitor for the cooling coil systems in Savannah River Site (SRS) waste tanks for over 40 years. Due to their age and operating history, cooling coils occasionally fail allowing chromate water to leak into the environment. When the leaks spill 10 lbs. or more of sodium chromate over a 24-hr period, the leak incidents are classified as Unusual Occurrences (UO) per CERCLA (Comprehensive Environmental Response, Compensation and Liability Act). The cost of reporting and cleaning up chromate spills prompted High Level Waste Engineering (HLWE) to initiate a study to investigate alternative tank cooling water inhibitor systems and the associated cost of replacement. Several inhibitor systems were investigated as potential alternatives to sodium chromate. All would have a lesser regulatory impact, if a spill occurred. However, the conversion cost is estimated to be $8.5 million over a period of 8 to 12 months to convert all 5 cooling systems. Although each of the alternative inhibitors examined is effective in preventing corrosion, there is no inhibitor identified that is as effective as chromate. Assuming 3 major leaks a year (the average over the past several years), the cost of maintaining the existing inhibitor was estimated at $0.5 million per year. Since there is no economic or regulatory incentive to replace the sodium chromate with an alternate inhibitor, HLWE recommends that sodium chromate continue to be used as the inhibitor for the waste tank cooling systems

  3. Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

    2016-01-15

    Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

  4. Development of LEAP-JET code for sodium-water reaction analysis. Validation by sodium-water reaction tests (SWAT-1R)

    International Nuclear Information System (INIS)

    Seino, Hiroshi; Hamada, Hirotsugu

    2004-03-01

    The sodium-water reaction event in an FBR steam generator (SG) has influence on the safety, economical efficiency, etc. of the plant, so that the selection of design base leak (DBL) of the SG is considered as one of the important matters. The clarification of the sodium-water reaction phenomenon and the development of an analysis model are necessary to estimate the sodium-water reaction event with high accuracy and rationality in selecting the DBL. The reaction jet model is pointed out as a part of the necessary improvements to evaluate the overheating tube rupture of large SGs, since the behavior of overheating tube rupture is largely affected by the reaction jet conditions outside the tube. Therefore, LEAP-JET has been developed as an analysis code for the simulation of sodium-water reactions. This document shows the validation of the LEAP-JET code by the Sodium-Water Reaction Test (SWAT-1R). The following results have been obtained: (1) The reaction rate constant, K, is estimated at between 0.001≤K≤0.1 from the LEAP-JET analysis of the SWAT-1R data. (2) The analytical results on the high-temperature region and the behaviors of reaction consumption (Na, H 2 O) and products (H 2 , NaOH, Na 2 O) are considered to be physically reasonable. (3) The LEAP-JET analysis shows the tendency of overestimation in the maximum temperature and temperature distribution of the reaction jet. (4) In the LEAP-JET analysis, the numerical calculation becomes unstably, especially in the mesh containing quite small sodium mass. Therefore, it is necessary to modify the computational algorism to stabilize it and obtain the optimum value of K in sodium-water reactions. (author)

  5. A statistical analysis on failure-to open/close probability of pneumatic valve in sodium cooling systems

    International Nuclear Information System (INIS)

    Kurisaka, Kenichi

    1999-11-01

    The objective of this study is to develop fundamental data for examination on efficiency of preventive maintenance and surveillance test from the standpoint of failure probability. In this study, as a major standby component, a pneumatic valve in sodium cooling systems was selected. A statistical analysis was made about a trend of valve in sodium cooling systems was selected. A statistical analysis was made about a trend of valve failure-to-open/close (FTOC) probability depending on number of demands ('n'), time since installation ('t') and standby time since last open/close action ('T'). The analysis is based on the field data of operating- and failure-experiences stored in the Component Reliability Database and Statistical Analysis System for LMFBR's (CORDS). In the analysis, the FTOC probability ('P') was expressed as follows: P=1-exp{-C-En-F/n-λT-aT(t-T/2)-AT 2 /2}. The functional parameters, 'C', 'E', 'F', 'λ', 'a' and 'A', were estimated with the maximum likelihood estimation method. As a result, the FTOC probability is almost expressed with the failure probability being derived from the failure rate under assumption of the Poisson distribution only when valve cycle (i.e. open-close-open cycle) exceeds about 100 days. When the valve cycle is shorter than about 100 days, the FTOC probability can be adequately estimated with the parameter model proposed in this study. The results obtained from this study may make it possible to derive an adequate frequency of surveillance test for a given target of the FTOC probability. (author)

  6. Radio-contaminant behaviour in the cover-gas space and the containment building of a sodium-cooled fast reactor in accident conditions

    International Nuclear Information System (INIS)

    Mathe, Emmanuel

    2014-01-01

    In the context of the Generation IV initiative, the consequences of a severe-accident (SA) in a sodium-cooled fast reactor must be studied. A SFR (Sodium cooled Fast Reactor) severe accident involves the disruption of the core by super-criticality involving the destruction of a certain number of fuel assemblies. Subsequently the interaction between hot fuel and liquid sodium can lead to a vapor explosion which could create a breach in the primary system. Some contaminated liquid sodium would thus be ejected into the containment building. In this situation, the evaluation of potential releases to the environment (the source term) must forecast the quantity and the chemical speciation of the radio-contaminants likely to be released from the containment building. One critical risk of a SA is the production of contaminated aerosols in the containment building by spray ejection of primary-system sodium. Being pyrophoric, the sodium droplets react with oxygen first oxidizing then burning, with significant heat of combustion. As well as evaluating the consequences of a pressure rise inside the containment, the evolution of the sodium must be assessed since not only is it activated and contaminated but, in oxide form, very toxic. Ultimately, the aerosols are the main radiological risk acting as the vector for radionuclide transport to the environment in the event of a problem with the confinement. These aerosols could evolve and interact with the FP (Fissile Products) and these interactions could modify the physical and chemical nature of the PF. We model a large part of the events that occur during a SA inside a SFR from the sodium spray fire to the reaction between sodium aerosols and PF (iodine). At first, we develop a numerical model (NATRAC) that simulates the sodium spray fire, calculates the temperature and the pressure inside the containment as well as the mass of aerosols produced during this kind of fire. The simulation has been validated with different

  7. A neutronics study for improving the safety and performance parameters of a 3600 MWth Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Sun, Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Chawla, Rakesh

    2013-01-01

    Highlights: ► The potential for neutronics design optimization is assessed for a large SFR core. ► Both beginning-of-life and equilibrium fuel cycle conditions are considered. ► The sodium void effect is decomposed via a neutron balance based methodology. ► The optimized core options adopt an appropriate sodium plenum design to reduce the void effect. ► The introduction of moderator pins is considered for enhancing the Doppler effect. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many performance advantages, but has one dominating neutronics drawback – a positive sodium void reactivity. The starting point for the present study is an SFR core design considered in the Collaborative Project on the European Sodium-cooled Fast Reactor (CP-ESFR). The aim is to analyze, for this reference core, four safety and performance parameters from the viewpoint of four different optimization options, and to propose possible optimized core designs. In doing so, the study focuses not only on the beginning-of-life state of the core, but also on the beginning of equilibrium closed fuel cycle. The four studied optimization options are: (a) introducing an upper sodium plenum and boron layer, (b) varying the core height-to-diameter (H/D) ratio, (c) introducing moderator pins into the fuel assembly, and (d) modifying the initial plutonium content. The sensitivity of the void reactivity, Doppler constant, nominal reactivity and breeding gain has been evaluated. In particular, the void reactivity, which is the most crucial safety parameter for the SFR, has been decomposed into its reaction-wise, isotope-wise and energy-group-wise components using a methodology based on the neutron balance equation. Extended voiding in the upper sodium plenum region – in conjunction with the effect of a boron layer introduced above the plenum – is found to be particularly effective in the void effect reduction while, at the same time

  8. Computation, measurement and analysis of the reactivity-to-power-transfer-function for the sodium cooled nuclear power plant KNK I

    International Nuclear Information System (INIS)

    Hoppe, P.; Mitzel, F.

    1977-02-01

    The Reactivity-to-Power-Transfer-Function for the sodium cooled nuclear power plant KNK I (Kompakte Natriumgekuehlte Kernenergieanlage) has been measured and compared with theoretical results. The measurements have been performed with the help of pseudostochastic reactivity perturbations. The transfer function has been determined by computing the auto- and cross-power-spectral-densities for the reactivity- and neutron flux signals. The agreement between the experimental and theoretical transfer function could be improved by adjusting the reactivity coefficients. The applications of these measurements with respect to reactor diagnosis and malfunction detection are discussed. For this purpose the accuracy of the measured transfer function is of great importance. Therefore an extensive error analysis has been performed. It turned out, that the inherent instability of the reactor without control system and the feedback by the primary coolant system were the reasons for comparatively big systematical errors. The conditions have been derived under which these types of errors can be considerably reduced. The conclusions can also be applied to analogical measurements at fast sodium cooled reactors. Because of their inherent stability the systematical errors will be reduced. (orig.) [de

  9. A study on the design method of the laminated rubber bearing for FBR

    International Nuclear Information System (INIS)

    Mazda, T.; Ishida, K.; Ikeuchi, K.; Yashiro, T.

    1994-01-01

    The Central Research Institute of Electric Power Industry (Japan) has been carrying out the Demonstration Test and Research Program of the Seismic Isolation System for the Fast Breeder Reactor (FBR) under contract with the Ministry of International Trade and Industry (FY1987--FY1993). In this research program, development of the seismic isolation element, reliability evaluation of the seismic isolation system, and examination of a design-based earthquake have been conducted in order to develop and propose design guidelines of seismic isolation system for FBR plant. The purpose of this paper is to describe the test results concerned with structural design of the laminated rubber bearing, and to explain the construction of the design method for the laminated rubber bearing, which is one of the most important factors in the design guidelines of the seismic isolation system

  10. The experimental sodium facility NAVA

    International Nuclear Information System (INIS)

    Langenbrunner, H.; Grunwald, G.; May, R.

    1976-01-01

    Within the framework of preparations for the introduction of sodium cooled fast breeder reactors an experimental sodium facility was installed at the Central Institute of Nuclear Research at Rossendorf. Design, engineering aspects and operation of this facility are described; operating experience is briefly discussed. (author)

  11. FFTF sodium and cover gas characterization and purification

    International Nuclear Information System (INIS)

    McCown, J.J.; Bloom, G.R.; Meadows, G.E.; Mettler, G.W.

    1980-02-01

    The FFTF Primary and Secondary Heat Transport System (HTS) sodium is purified with cold traps which have packed crystallizers and external economizers. The Primary HTS cold trap is NaK cooled and the Secondary HTS cold traps are air cooled. The FFTF cold traps have maintained high purity in the sodium since sodium fill. Plant operational procedures during fill and initial sodium heatup to 800 0 F were controlled to assure low release rates of impurities to the sodium. The FFTF sodium systems are monitored by plugging temperature indicators and by several sampling methods. During reactor fill and non-fueled operations at 400 to 800 0 F, impurity changes in the sodium were followed by continuous plugging indicator coverage, by exposing wires and foils to measure carbon, hydrogen and oxygen, and by bulk sample analysis of all other trace constituents. The sampling and analysis methods and data are presented, impurity excursions in the cover gas and sodium are described, and impurity trends are discussed

  12. Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

  13. Mechanical properties of structural materials for FBR sodium application. Semi-annual progress report for period ending January 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Metallographic evaluations of the CRBR core barrel forging material, creep rupture tested at 538/sup 0/C in air, were performed. The majority of the specimens had a knobby appearance on the surface of the gage section. The stress-rupture life for sodium pre-exposed Type 316 stainless steel performed at 538/sup 0/C in flowing sodium is increased by a factor of at least three at a stress of 275.8 MPa (40.0 ksi) when compared to tests in sodium for as-received material (mill annealed) at the same conditions. Creep-rupture tests of mill annealed type 316 stainless steel in flowing sodium at 593/sup 0/C and 224.1 MPa (32.5 ksi), involving different gage diameters of 0.25, 0.15, and 0.10 inches, were evaluated. A creep-rupture test of an alloy 718 specimen tested at 649/sup 0/C and 344.7 MPa (50.0 ksi) in the flowing sodium, after exposure to flowing sodium at 649/sup 0/C for 10,000 hours, ruptured after 9617 hours. It is estimated that after nearly 20,000 hours in sodium, the rupture life was reduced approximately 30% when compared to results for as-received material tested in flowing sodium (and air).

  14. Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties. (author)

  15. Development of extreme rainfall PRA methodology for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

    2016-01-01

    The objective of this study is to develop a probabilistic risk assessment (PRA) methodology for extreme rainfall with focusing on decay heat removal system of a sodium-cooled fast reactor. For the extreme rainfall, annual excess probability depending on the hazard intensity was statistically estimated based on meteorological data. To identify core damage sequence, event trees were developed by assuming scenarios that structures, systems and components (SSCs) important to safety are flooded with rainwater coming into the buildings through gaps in the doors and the SSCs fail when the level of rainwater on the ground or on the roof of the building becomes higher than thresholds of doors on first floor or on the roof during the rainfall. To estimate the failure probability of the SSCs, the level of water rise was estimated by comparing the difference between precipitation and drainage capacity. By combining annual excess probability and the failure probability of SSCs, the event trees led to quantification of core damage frequency, and therefore the PRA methodology for rainfall was developed. (author)

  16. Coolant-fuel interaction in Sodium-cooled Fast Reactors: Structural investigations of The Na-An-O (An = U, Np, Pu) systems

    International Nuclear Information System (INIS)

    Smith, A.L.; Raison, P.E.; Bykov, D.M.; Konings, R.J.; Caciuffo, R.; Cheetham, A.K.

    2014-01-01

    Nuclear energy has the potential to provide Europe with a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gases emissions. The interest for Sodium-cooled-Fast-spectrum Reactors (SFRs), when compared to Pressurized Water Reactors (PWRs), lies in their more efficient management of plutonium and other actinides as well as their ability to use almost all of the energy in the natural uranium versus 1% utilized in thermal spectrum systems. The high fuel efficiency of fast reactors could greatly dampen concerns about fuel supply. But these reactors have also several drawbacks when compared to PWRs (i.e sodium fire, Na reaction with O2 and H2O, interaction of sodium with oxide fuels). Their development at an industrial scale needs therefore an exhaustive safety assessment that comprises both experimental work and development of sophisticated modelling tools able to describe the reactor behaviour in normal or incidental conditions

  17. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ponciroli, R.; Passerini, S.; Vilim, R. B.

    2016-04-17

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

  18. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, EC-JRC, Westerduinweg 3, P.O. Box 2, NL-0 1755 ZG Petten (Netherlands)

    2006-07-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  19. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    International Nuclear Information System (INIS)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-01-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  20. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (3) Progress of component design

    International Nuclear Information System (INIS)

    Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi; Eto, Masao; Miyagawa, Takayuki

    2015-01-01

    In the frame work of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors have been developing in the circumstance of worldwide deployment of SFRs. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC in the feasibility study of SDG for Sodium-cooled Fast Reactor (SFR). In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX (pump/IHX) was modified for the primary heat exchanger (PHX), which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator (SG), protective wall tube type design is under investigation as an option with less R and D risks. (author)

  1. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-01

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean

  2. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  3. Some difference of concepts between design guideline for FBR base isolation system and aseismic design guideline of LWR in Japan

    International Nuclear Information System (INIS)

    Shibata, Heki

    1992-01-01

    This paper deals with the concept and the relation of 'the Base Isolation System and FBR' to the Safety Criteria and the Guideline of the Aseismic Design of LWR in Japan. The Central Research Institute of Electric Power Industries have been working for FBR last several years. The author has been contribute to their works, and this is one of the subjects. He described his own idea obtained through the cooperative work with CRIEPI. (author)

  4. Recoverying device for sodium vapor in inert gas

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tamotsu; Nagashima, Ikuo

    1992-11-06

    A multi-pipe type heat exchanger for cooling an inert gas and a mist trap connected to the inert gas exit of the heat exchanger are disposed. A mist filter having bottomed pipes made of an inert gas-permeable sintered metal is disposed in the mist trap, and an inert gas discharge port is disposed at the upper side wall. With such a constitution, a high temperature inert gas containing sodium vapors can be cooled efficiently by the multi-pipe type heat exchanger capable of easy temperature control, thereby converting sodium vapors into mists, and the inert gas containing sodium mists can be flown into the mist trap. Sodium mists are collected by the mist filter and sodium mists flown down are discharged from the discharge port. With such procedures, a great amount of the inert gas containing sodium vapors can be processed continuously. (T.M.).

  5. 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)

  6. Resistance of Alkali Activated Water-Cooled Slag Geopolymer to Sulphate Attack

    Directory of Open Access Journals (Sweden)

    S. A. Hasanein

    2011-06-01

    Full Text Available Ground granulated blast furnace slag is a finely ground, rapidly chilled aluminosilicate melt material that is separated from molten iron in the blast furnace as a by-product. Rapid cooling results in an amorphous or a glassy phase known as GGBFS or water cooled slag (WCS. Alkaline activation of latent hydraulic WCS by sodium hydroxide and/or sodium silicate in different ratios was studied. Curing was performed under 100 % relative humidity and at a temperature of 38°C. The results showed that mixing of both sodium hydroxide and sodium silicate in ratio of 3:3 wt.,% is the optimum one giving better mechanical as well as microstructural characteristics as compared with cement mortar that has various cement content (cement : sand were 1:3 and 1:2. Durability of the water cooled slag in 5 % MgSO4 as revealed by better microstructure and high resistivity-clarifying that activation by 3:3 sodium hydroxide and sodium silicate, respectively is better than using 2 and 6 % of sodium hydroxide.

  7. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

    2015-01-01

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  8. Assessment of flow induced vibration in a sodium-sodium heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Prakash, V. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)], E-mail: prakash@igcar.gov.in; Thirumalai, M.; Prabhakar, R.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

    2009-01-15

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam. It is a liquid metal sodium cooled pool type fast reactor with all primary components located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to the secondary sodium in a sodium to sodium Intermediate Heat Exchanger (IHX), which in turn is transferred to water in the steam generator. PFBR IHX is a shell and tube type heat exchanger with primary sodium on shell side and secondary sodium in the tube side. Since IHX is one of the critical components placed inside the radioactive primary sodium, trouble-free operation of the IHX is very much essential for power plant availability. To validate the design and the adequacy of the support system provided for the IHX, flow induced vibration (FIV) experiments were carried out in a water test loop on a 60 deg. sector model. This paper discusses the flow induced vibration measurements carried out in 60 deg. sector model of IHX, the modeling criteria, the results and conclusion.

  9. Sodium fires and nuclear power station safety

    International Nuclear Information System (INIS)

    Ivanenko, V.N.; Zubin, A.; Drobyshev, A.V.

    1986-01-01

    The danger of sodium aerosol release at a design basis accident (DBA) of a sodium-cooled fast reactor that involves coolant leakage and burning, is being analyzed. It has been shown that radioactive and toxic releases at DBA do not exceed permissible values. Some means of sodium fire fighting are described. (author)

  10. Evaluation of core distortion in FBR

    International Nuclear Information System (INIS)

    Ikarimoto, I.; Tanaka, M.; Okubo, Y.

    1984-01-01

    The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

  11. Water Mock-up for the Sodium Waste Treatment Process

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Kim, Jong Man; Kim, Byung Ho; Lee, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    It is important to safely treat the waste sodium which was produced from the sodium cooled fast reactors and the sodium facilities. About 1.3 tons of sodium waste has accumulated at KAERI from the sodium experiments which have been carried out since 1990. Also, large scaled sodium experiments are scheduled to verify the design of the sodium cooled fast reactor. As a treatment method for the waste sodium produced at the sodium facility, an investigation of the reaction procedure of the waste sodium with the sodium hydroxide aqueous has been developed. The NOAH process was developed in France for the treatment of waste sodium produced from sodium facilities and reactors. In the NOAH process, a small amount of sodium waste is continuously injected into the upper space which is formed on the free surface of the aqueous and slowly reacted with sodium hydroxide aqueous. Since the density of the sodium is lower than that of the aqueous, the injected sodium waste sometimes accumulates above the free surface of the sodium hydroxide aqueous, and its reaction rate becomes slow or suddenly increases. In the improved process, the sodium was injected into a reaction vessel filled with a sodium hydroxide aqueous through an atomizing nozzle installed on a lower level than that of the aqueous to maintain the reaction uniformly. Fig.1 shows the sodium waste process which was proposed in KAERI. The aqueous is composed of 60% sodium hydroxide, and its temperature is about 60 .deg. C. The process is an exothermic reaction. The hydrogen gas is generated, and the concentration of the sodium hydroxide increases in this process. It needs several systems for the process, i.e. a waste sodium injection, a cooling of the aqueous, hydrogen ventilation, and neutralization with nitric acid. The atomizing nozzle was designed to inject the sodium with the nitrogen gas which supplies a heat to the sodium to prevent its solidification and to uniformly mix the sodium with the aqueous. There are

  12. Recent progress in sodium technology

    Energy Technology Data Exchange (ETDEWEB)

    Hallett, W. J.

    1963-10-15

    Progress over the past year in U. S. laboratories studying some of the materials and engineering problems that must be resolved in bringing the technology of sodium to an economically and technically attractive point is reviewed. The status of sodium cooled power reactors in the U. S. is described. (P.C.H.)

  13. Conceptual Design for BOP of the Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoo, Tae Geun; Kim, Seong O; Kim, Eui Kwang; Seong, Seung Hwan

    2010-01-01

    The heavy dependence on nuclear power eventually raise the issues of an efficient utilization of uranium resources, which Korea presently imports from abroad, end of a spent fuel storage. From the viewpoint that sodium-cooled fast Reactors (SFR s ) have the potential of an enhanced safety by utilizing inherent safety characteristics, trans-uranics (TRU) reduction and resolving the spent fuel storage problems through a proliferation-resistant actinide recycling. SFR s are sure to be most promising nuclear power operation. The Korea Atomic Energy Research Institute (KAERI) has been developing SFR design technologies since 1997. And nowadays, the preliminary heat balance of the demonstration SFR is calculated. However, in order to verify design condition of the NSSS, it is necessary to set the heat balance and the conceptual design for BOP of the SFR as a part of the SFR design technique development business. Moreover, in order to confirm whether the heat balance can actually appropriate via the turbine characteristic, it is required to carry out the performance analysis of the turbine cycle. For that, the main purposes of this study are; 1) to derivate the conceptual design for BOP, 2) to analyze the performance of the turbine cycle, 3) to derivate the main consideration for BOP design

  14. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  15. Study on reprocessing plant during transition period from LWR to FBR

    International Nuclear Information System (INIS)

    Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

    2011-01-01

    We have proposed a concept of a reprocessing plant suitable for the transition period from the light water reactors (LWRs) to the fast breeder reactors (FBRs) by making comparison of two plant concepts: (1) Independent Plant which processes LWR fuel and FBR fuel in separately constructed lines and (2) Modularized Plant which processes LWR fuel and FBR fuel in a same line. We made construction plans based on the reference power generation plan, and evaluated the Pu supply capability using the power generation plan as an indicator of plant operation flexibility. In general, a margin of processing capacity increases the Pu supply capability. The margin of the Modularized Plant necessary to obtain equivalent Pu supply capability is smaller than that of the Independent Plant. Also the margin of the Independent Plant results in decrease in the plant utilization factor. But the margin of the Modularized Plant results in little decrease in the plant utilization factor, because the Modularized Plant can address the types of reprocessing fuel to adjust to Pu demand and processing capacity. Therefore, the Modularized Plant has a greater potential for the reprocessing plants during transition period. (author)

  16. Two neural network based strategies for the detection of a total instantaneous blockage of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Martinez-Martinez, Sinuhe; Messai, Nadhir; Jeannot, Jean-Philippe; Nuzillard, Danielle

    2015-01-01

    The total instantaneous blockage (TIB) of an assembly in the core of a sodium-cooled fast reactor (SFR) is investigated. Such incident could appear as an abnormal rise in temperature on the assemblies neighbouring the blockage. Its detection relies on a dataset of temperature measurements of the assemblies making up the core of the French Phenix Nuclear Reactor. The data are provided by the French Commission of Atomic and Alternatives Energies (CEA). Here, two strategies are proposed depending on whether the sensor measurement of the suspected assembly is reliable or not. The proposed methodology implements a time-lagged feed-forward neural (TLFFN) Network in order to predict the one-step-ahead temperature of a given assembly. The incident is declared if the difference between the predicted process and the actual one exceeds a threshold. In these simulated conditions, the method is efficient to detect small gradients as expected in reality. - Highlights: • We study the total instantaneous blockage (TIB) of a sodium-cooled fast reactor. • The TIB symptom is simulated as an abrupt rise on temperature (0.1–1 °C/s). • The goal is to improve the early detection of the incident. • Two strategies laying on neural networks are proposed. • TIB is detected in 3 s for 1 °C/s and 18–21 s for 0.1 °C/s

  17. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Sun Kaichao, E-mail: kaichao.sun@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland); Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland)

    2011-07-15

    Highlights: > We analyze the void reactivity effect for three ESFR core fuel cycle states. > The void reactivity effect is decomposed by neutron balance method. > Novelly, the normalization to the integral flux in the active core is applied. > The decomposition is compared with the perturbation theory based results. > The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly by the

  18. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sun Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro; Chawla, Rakesh

    2011-01-01

    Highlights: → We analyze the void reactivity effect for three ESFR core fuel cycle states. → The void reactivity effect is decomposed by neutron balance method. → Novelly, the normalization to the integral flux in the active core is applied. → The decomposition is compared with the perturbation theory based results. → The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly

  19. Study on commercial FBR concepts by combining innovative technologies

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.; Kuroha, M.; Hida, T.

    1992-01-01

    A study was conducted on future prospects of FBR commercialization. Targets of further improving safety and economy were set to make commercial power plants that would be superior to future LWRs. Promising innovative technologies studied domestically and overseas were extracted by evaluating prospects for commercialization, effect, and plant applicability. Several commercial plants were conceptualized by introducing such technology to large-scale and oxide-fuel reactors. Estimates of construction cost, etc., proved that the targets could be achieved. A concept of long-term technological development was synthesized. (author)

  20. Method of detecting coolant leakages from the pipeways in FBR type reactors

    International Nuclear Information System (INIS)

    Sonoda, Yukio; Tamaoki, Tetsuo

    1986-01-01

    Purpose: To detect coolant leakage in the incore pipeways of loop type FBR type reactors in the initial stage at high sensitivity. Constitution: Temperature of the coolants sealed between incore pipeways and the buffle surrounding them is measured by thermocouples and coolant leakage is detected due to fluctuating components. A well-insertion type in which electrode is sealed with argon is used as the thermo-couples. Signals from the thermocouples are once amplified, removed with DC components and then only the fluctuating components are outputted. The fluctuating components are digitalized, passed through an adaptive digital filter and the RMS value as the difference between the output signal and the thermocouple signal is calculated. The calculated value is compared with a threshold value in a comparative calculator. If it exceeds the threshold value, it is judged as abnormal to display an alarm on an alarm display. In this way, the coolant leakage for the pipeways in the FBR type reactor can be detected on real time and at high sensitivity. (Kamimura, M.)

  1. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  2. Study on in-service inspection and repair program and related plant design for Japan Sodium-Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Suzuki, Shinichi; Kotake, Shoji; Nishiyama, Noboru; Uzawa, Masayuki

    2011-01-01

    Maintenance and repair program and conformity with them were investigated as a part of the conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR). The maintenance program was set by taking the feature of sodium-cooled reactors and domestic practice of LWRs into account. Both of regulatory required inspection and voluntary inspection, which are conducted in the domestic LWRs, were counted. The regulatory required ISI program was based on that of the previous Japanese SFRs, LWRs (JSME S NA1) and liquid metal cooled reactors (ASME section XI division 3). Parts to be inspected, methods of inspection were identified for major structures and components. Concerning the repair program, we set three levels of repair requirements based on estimated frequency of defect and failure during the plant life time. For level 1, which might be occur several times during the plant life time, it is required to be easily repaired in a short period. Access routes and working space are considered in the component design and its arrangement. For level 2, which might be unlikely to occur during the plant life time, it is required to check that the repair work is feasible in a practical time range. For level 3, which frequency is negligible small, repair is not taken into account but the feasibility was investigated. The plant design shall be done so that all of above mentioned inspection and repair can be conducted. It is desired to ensure accessibility for all of the coolant and cover gas boundaries and the internal structures in order to cope with unforeseen troubles. Access routes for the reactor vessel and its internal structures, piping, pumps and intermediate heat exchangers and steam generators were investigated. As the results of that, possible ways for implementation of the maintenance and repair were identified. (author)

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

    International Nuclear Information System (INIS)

    Abramson, R.

    1984-04-01

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

  4. Three-dimensional tsunami analysis for the plot plan of a sodium-cooled fast reactor plant

    International Nuclear Information System (INIS)

    Hayakawa, Satoshi; Watanabe, Osamu; Itoh, Kei; Yamamoto, Tomohiko

    2013-01-01

    As the practical evaluation method of the effect of tsunami on buildings, the formula of tsunami force has been used. However, it cannot be applied to complex geometry of buildings. In this study, to analyze the effect of tsunami on the buildings of sodium-cooled fast reactor plant more accurately, three-dimensional tsunami analysis was performed. In the analysis, VOF (Volume of Fluid) method was used to capture free surface of tsunami. At the beginning, it was confirmed that the tsunami experiment results was reproduced by VOF method accurately. Next, the three-dimensional tsunami analysis was performed with VOF method to evaluate the flow field around the buildings of the plant from the beginning of the tsunami until the backwash of that. (author)

  5. Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chenu, A.

    2011-10-01

    Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

  6. Evaluation of Microstructural and Mechanical Property of Medium-sized HT9 Cladding Forged Material for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Lee, Kang Soo; Kim, Sung Ho; Lee, Chan Bock

    2012-01-01

    Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at 1170 degrees C after the induction melting to make round bar as 160 mm diameter, 7000 mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2-3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120 mm.

  7. Minor actinides transmutation potential: state of art for GEN IV sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Buiron, Laurent

    2015-01-01

    In the frame of the R and D program relative to the 1991 French act on nuclear waste management, fast neutron systems have shown relevant characteristics that meet both requirements on sustainable resources management and waste minimization. They also offer flexibility by mean of burner or breeder configurations allowing mastering plutonium inventory without significant impact on core safety. From the technological point of view, sodium cooled fast reactor are considered in order to achieve mean term industrial deployment. The present document summaries the main results of R and D program on minor actinides transmutation in sodium fast reactor since 2006 following recommendation of the first part of the 1991 French act. Both homogeneous and heterogeneous management achievable performances are presented for 'evolutionary' SFR V2B core as well as low void worth CFV core for industrial scale configurations (1500 MWe). Minor actinides transmutation could be demonstrated in the ASTRID reactor with the following configurations: - a 2%vol Americium content for the homogeneous mode, - a 10%vol Americium content for the heterogeneous mode, without any substantial modification of the main core safety parameters and only limited impacts on the associated fuel cycle (manufacturing issues are not considered here). In order to achieve such goal, a wide range of experimental irradiations driven by transmutation scenarios have to be performed for both homogeneous and heterogeneous minor actinides management. (author) [fr

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

    International Nuclear Information System (INIS)

    2014-11-01

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

  9. 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.)

  10. The influence of sodium fires on LMFBRs safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Justin, F [DSN/Centre de Fontenay-aux-Roses, Fontenay-aux-Roses (France)

    1979-03-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs.

  11. The influence of sodium fires on LMFBRs safety analysis

    International Nuclear Information System (INIS)

    Justin, F.

    1979-01-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs

  12. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  13. Development of end plug welding method in the fabrication of FBR fuel pins

    International Nuclear Information System (INIS)

    Ohtani, Seiji; Sawayama, Takeo; Tateishi, Yoshinori

    1977-01-01

    As a part of the development of the automatic and remote controlled fabrication of FBR fuel pins, welding of fuel pin end plugs has been examined. Cladding tubes and end plugs used for this experiment are made of SUS 316, and they are the components of fuel pins for the prototype fast breeder reactor (Monju) or the second core of Joyo (Joyo MK-II). The welding tests of cladding tubes and four kinds of end plugs were carried out by means of two techniques; tungsten inert gas welding and laser welding. It can be said that no considerable difference was observed in weld penetration, occurrence rate of weld defects and breaking strength between the tight fit and the loose fit plugs. The face-to-face fit welding requires the least welding heat input, but involves much difficulty in the control of weld penetration and bead zone diameter. The good concentrative property and high energy density of laser beam make the face of weld hollow due to the vaporization of weld metal. However, this problem can be easily solved by changing the shape of end plugs. Good results in the other characteristics of the weld also were obtained by this laser welding. Further experiment is needed in connection with the compatibility of weld metal with sodium and neutron irradiation before final judgement is made on the laser welding technique. (Nakai, Y.)

  14. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    International Nuclear Information System (INIS)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

    2008-01-01

    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities

  15. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

    2008-01-15

    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

  16. Characterization of the liquid sodium spray generated by a pipework hole

    International Nuclear Information System (INIS)

    Torsello, G.; Parozzi, F.; Nericcio, L.; Araneo, L.; Cozzi, F.; Carcassi, M.; Mattei, N.

    2012-01-01

    Due to its advantageous thermodynamic characteristics at high temperature (550 deg. C), liquid sodium is the main candidate to be the cooling fluid for Generation TV nuclear reactors SFR (Sodium-cooled Fast Reactors). Now, sodium reacts very violently, both with the water and the oxygen of the air. Only few data were known about the liquid sodium behaviour when spread in the environment through micro defects. These are often present in a cooling circuit in welded or sealed joints and more rarely in the pipes. Micro defects, on the other hand, can be also generated in a cooling circuit because of the vibrations always present in a circuit into which a fluid runs. A new set-up, named LISOF, was built for testing high temperature liquid sodium when passing through micro defects and generating sprays or jets. Sprays and jets were generated by means of nozzles embedding sub milli-metric holes the diameter of which was: 0.2 mm, 0.4 mm, 0.5 mm. Tests were performed by pressurizing liquid sodium (550 deg. C) at: 3, 6 and 9 barg. Normal and high speed cinematography were used for the direct observation of the liquid sodium sprays while Phase Doppler Interferometry was used for the measurement of the droplets characteristics and velocity. Tests concerning the behaviour of the high temperature liquid sodium firing in air or in contact with the cement cover applied to a scaled down core catcher simulacrum were also performed. The paper presents the built set-up and the collected results. (authors)

  17. Characterization of the liquid sodium spray generated by a pipework hole

    Energy Technology Data Exchange (ETDEWEB)

    Torsello, G.; Parozzi, F.; Nericcio, L. [RSE - Nuclear and Industrial Plant Safety Team, Power Generation System Dept., via Rubattino 54, 20134 Milano (Italy); Araneo, L.; Cozzi, F. [Politecnico di Milano, Energy Dept., via Lambruschini 4, 20156 Milano (Italy); Carcassi, M.; Mattei, N. [Universita di Pisa-Facolta d' Ingegneria DIMNP-Mechanical, Nuclear and Production Dep., Largo L. Lazzarino 2, 56126 Pisa (Italy)

    2012-07-01

    Due to its advantageous thermodynamic characteristics at high temperature (550 deg. C), liquid sodium is the main candidate to be the cooling fluid for Generation TV nuclear reactors SFR (Sodium-cooled Fast Reactors). Now, sodium reacts very violently, both with the water and the oxygen of the air. Only few data were known about the liquid sodium behaviour when spread in the environment through micro defects. These are often present in a cooling circuit in welded or sealed joints and more rarely in the pipes. Micro defects, on the other hand, can be also generated in a cooling circuit because of the vibrations always present in a circuit into which a fluid runs. A new set-up, named LISOF, was built for testing high temperature liquid sodium when passing through micro defects and generating sprays or jets. Sprays and jets were generated by means of nozzles embedding sub milli-metric holes the diameter of which was: 0.2 mm, 0.4 mm, 0.5 mm. Tests were performed by pressurizing liquid sodium (550 deg. C) at: 3, 6 and 9 barg. Normal and high speed cinematography were used for the direct observation of the liquid sodium sprays while Phase Doppler Interferometry was used for the measurement of the droplets characteristics and velocity. Tests concerning the behaviour of the high temperature liquid sodium firing in air or in contact with the cement cover applied to a scaled down core catcher simulacrum were also performed. The paper presents the built set-up and the collected results. (authors)

  18. Pu Denaturing by Transmutation of MA in FBR Multi-cycle

    Energy Technology Data Exchange (ETDEWEB)

    Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi [Tokyo Institute of Technology, 2-12-1-N1-1 Ookayama, Meguro-ku, Tokyo, 1528550 (Japan)

    2009-06-15

    Pu accumulation and its recycling is important in the term of energy resources, however one of the most sensitive issues is non-proliferation in the future fuel cycle based on fast breeder reactor (FBR). The present paper utilizes Protected Pu Production (P{sup 3}) concept for the production of {sup 238}Pu and {sup 242}Pu by Minor Actinides (MA) transmutation to enhance the proliferation resistance of Pu in the fuel. Increase in the {sup 238}Pu and {sup 242}Pu isotopic fraction creates a high rate of internal heat generation by alpha decay (DH) and/or a high neutron source of spontaneous fission (SFN) in Pu that would be encountered during manufacturing and maintaining of nuclear explosive device. The feasibility of denaturing of Pu by MA transmutation in medium size FBR has been studied from the viewpoint of even-mass number Pu accumulation during multi-cycle of Pu and MA. The proliferation resistance property of Pu is also evaluated based on the specific decay heat and spontaneous fission neutron, compared with the reference criteria. In present paper, the P{sup 3} technology based on multi-recycled Pu and MA is compared with the conventional technology based on multi-recycled Pu only. The detail of mass balance behavior is, however, beyond the scope of the present paper. (authors)

  19. Feasibility study of application of ductless fuel assembly to FBR

    International Nuclear Information System (INIS)

    Itoh, K.; Shibahara, I.

    1996-01-01

    Feasibility studies on an application of the ductless fuel concept to an FBR core have been carried out in order to evaluate the basic features of the ductless core, especially in the fields of the thermal-hydraulic aspects and the mechanical behaviors. Regarding thermal-hydraulic aspects, analyses were performed by using a whole core thermal-hydraulic analysis code by making some modification for this study on the PWR code THINC. A small scaled ductless core model was prepared and a hydraulic experiment was carried out to study basic hydraulic characteristics of a ductless core. Core mechanical behaviors were analyzed focusing on the core irradiation bowing aspects and the seismic behaviors. Following features are revealed on the core structural behaviors: (1) the bowing stiffness of the ductless assembly is around 1/5 to 1/10 of that of the duct type assembly; (2) the contact loads between assemblies by the bowing effects are small through core cycles; (3) the damping of the ductless assemblies are so large that the seismic responses are small and the loads between assemblies are small due to occurring many contact points. Through this study it is expected that the concept of the ductless fuel can be applicable to FBR cores from the design view points of thermal-hydraulic and core mechanical behaviors

  20. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  1. LBB assessment on ferrite piping structure of large-scale FBR

    OpenAIRE

    兪 淵植

    2002-01-01

    These days, this interest on LBB(Leak before Break) design becomes to be rising in the viewpoint of the cost reduction and structural inter-grity for the commercialization of FBR plants, LBB design enables pla-nts to be shut down safely before occuring unstable fracture by dete- cting the leak rates even if a crack initiates and penetrates a wall thickness. It is necessary to assess crack growth and penetration be- havior considering in-service conditions under operation temperature, leak re...

  2. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  3. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  4. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

  5. An investigation of sodium iodide solubility in sodium-stainless steel systems

    International Nuclear Information System (INIS)

    Sagawa, Norihiko; Tashiro, Suguru

    1996-01-01

    Sodium iodide and major constituents of stainless steel in sodium are determined by using the steel capsules to obtain a better understanding on contribution of the constituents to the apparent iodide solubility in sodium. The capsule loaded with 20 g sodium and 0.1 - 0.3 g powder of sodium iodide is heated at its upper part in a furnace and cooled at its bottom on brass plates to establish a large temperature gradient along the capsule tube. After a given period of equilibration, the iodide and constituents are fixed in solidified sodium by quick quenching of the capsules. Sodium samples are taken from the sectioned capsule tube and submitted to sodium dissolution by vaporized water for determination of the iodine and to vacuum distillation for determination of the metal elements. Iron and nickel concentrations are observed to be lower in the samples at higher iodine concentrations. Chromium and manganese concentrations are seen to be insensitive to the iodine concentrations. The observations can be interpreted by a model that sodium oxide combines with metal iodide in sodium to form a complex compound and with consideration that the compound will fall and deposit onto the bottom of the capsule by thermal diffusion. (author)

  6. Formation and Transformation Behavior of Sodium Dehydroacetate Hydrates

    Directory of Open Access Journals (Sweden)

    Xia Zhang

    2016-04-01

    Full Text Available The effect of various controlling factors on the polymorphic outcome of sodium dehydroacetate crystallization was investigated in this study. Cooling crystallization experiments of sodium dehydroacetate in water were conducted at different concentrations. The results revealed that the rate of supersaturation generation played a key role in the formation of the hydrates. At a high supersaturation generation rate, a new sodium dehydroacetate dihydrate needle form was obtained; on the contrary, a sodium dehydroacetate plate monohydrate was formed at a low supersaturation generation rate. Furthermore, the characterization and transformation behavior of these two hydrated forms were investigated with the combined use of microscopy, powder X-ray diffraction (PXRD, Raman spectroscopy, Fourier transform infrared (FTIR, thermal gravimetric analysis (TGA, scanning electron microscopy (SEM and dynamic vapor sorption (DVS. It was found that the new needle crystals were dihydrated and hollow, and they eventually transformed into sodium dehydroacetate monohydrate. In addition, the mechanism of formation of sodium dehydroacetate hydrates was discussed, and a process growth model of hollow crystals in cooling crystallization was proposed.

  7. Heat resistant/radiation resistant cable and incore structure test device for FBR type reactor

    International Nuclear Information System (INIS)

    Tanimoto, Hajime; Shiono, Takeo; Sato, Yoshimi; Ito, Kazumi; Sudo, Shigeaki; Saito, Shin-ichi; Mitsui, Hisayasu.

    1995-01-01

    A heat resistant/radiation resistant coaxial cable of the present invention comprises an insulation layer, an outer conductor and a protection cover in this order on an inner conductor, in which the insulation layer comprises thermoplastic polyimide. In the same manner, a heat resistant/radiation resistant power cable has an insulation layer comprising thermoplastic polyimide on a conductor, and is provided with a protection cover comprising braid of alamide fibers at the outer circumference of the insulation layer. An incore structure test device for an FBR type reactor comprises the heat resistant/radiation resistant coaxial cable and/or the power cable. The thermoplastic polyimide can be extrusion molded, and has excellent radiation resistant by the extrusion, as well as has high dielectric withstand voltage, good flexibility and electric characteristics at high temperature. The incore structure test device for the FBR type reactor of the present invention comprising such a cable has excellent reliability and durability. (T.M.)

  8. Effect of Reflector Material on the Neutronic Characteristics of the Small Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sung Hwan; Baek, Min Ho; Yoo, Jae Woon; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The sodium-cooled fast reactor (SFR) has been chosen as a candidate for the Gen-IV Nuclear Energy Systems Initiative due to the advantages in utilization of uranium resources and reduction of radioactive wastes. Recently, the uranium blanket concept is omitted for a purpose of the non-proliferation, hence the reflector material plays a more important role in reactor core design. Moreover, especially in the Korean prototype SFR, the initial core should startup with low-enriched uranium ({<=} 20 w/o) for 100 {approx} 150 MWe power. This restriction causes significant difficulties to achieve sufficient excess reactivity. Thus, in this paper, core characteristic studies of various reflector materials (HT9, BeO, MgO, and ZrH{sub 1.6}) are performed to enhance the initial core excess reactivity

  9. Storage and disposal of high-level radioactive waste from advanced FBR fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2011-01-01

    Waste management of fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P and T) technology was investigated by focusing on thermal constraints due to heat deposition from waste in storage and disposal facilities including economics aspects of those facilities. Partitioning of minor actinides (MAs) and heat-generating fission products in high-level waste can enlarge the containment ratio of waste elements in the glass waste forms and shorten predisposal storage period. Though MAs can be transmuted in FBRs or dedicated transmuters, heat-generating fission products are difficult to be transmuted; they are partitioned and stored for a long time before disposal. The disposal concepts for heat-generating fission products and remainders such as rare-earth elements depend on storage period that ranges from several years to several hundreds of years. Short-term storage results in small size of storage facilities and large size of repositories, and vice versa for long-term storage. This trade-off relation was analyzed by estimating repository size as a function of storage period. The result shows that transmutation of MAs is essentially effective to reduce repository size regardless to storage period, and a combination of P and T can provide a smaller repository than the conventional one by two orders of magnitude. The cost analysis for waste management was also made based on rough assumptions on storage, transportation and repository excluding cost for introducing P and T that are still under evaluation. Cost of waste management for FBR without P and T is 0.25 Yen/kWh that is slightly smaller than that for LWR without P and T, 0.30 Yen/kWh. The introduction of MA transmutation to the FBR results in cost of 0.20 Yen/kWh, and full introduction of P and T provides the smallest cost of 0.08 Yen/kWh. (author)

  10. Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1983-01-01

    This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

  11. Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

    International Nuclear Information System (INIS)

    Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

    2010-01-01

    A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

  12. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

    International Nuclear Information System (INIS)

    Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

    2009-01-01

    The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

  13. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  14. Method of preventing sodium from flowing when pipes of a fast breeder reactor are injured

    International Nuclear Information System (INIS)

    Nakai, Yasushi; Yamagishi, Yoshiaki; Koga, Tomonari.

    1975-01-01

    Object: To inject high pressure sodium into an inlet nozzle portion when fluid pressure in the inlet nozzle portion of a core cooling pipe on the inlet side is in an abnormal condition, to thereby quickly and positively prevent the flow of sodium in a high pressure chamber in a reactor vessel, when pipes are injured. Structure: When the core cooling pipe on the inlet side is injured and as a consequence the pressure gage detects an abnormal condition of fluid pressure in the inlet nozzle, the valve is opened to allow high pressure sodium to inject into the inlet nozzle through a high pressure sodium supply pipe, thereby blocking a back-flow of sodium in the high pressure chamber into the core cooling pipe. (Kamimura, M.)

  15. System design study of a membrane reforming hydrogen production plant using a small sized sodium cooled reactor

    International Nuclear Information System (INIS)

    Chikazawa, Y.; Konomura, M.; Hori, T.; Sato, H.; Uchida, S.

    2004-01-01

    In this study, a membrane reforming hydrogen production plant using a small sized sodium cooled reactor was designed as one of promising concepts. In the membrane reformer, methane and steam are reformed into carbon dioxide and hydrogen with sodium heat at a temperature 500 deg-C. In the equilibrium condition, steam reforming proceeds with catalyst at a temperature more than 800 deg-C. Using membrane reformers, the steam reforming temperature can be decreased from 800 to 500 deg-C because the hydrogen separation membrane removes hydrogen selectively from catalyst area and the partial pressure of hydrogen is kept much lower than equilibrium condition. In this study, a hydrogen and electric co-production plant has been designed. The reactor thermal output is 375 MW and 25% of the thermal output is used for hydrogen production (70000 Nm 3 /h). The hydrogen production cost is estimated to 21 yen/Nm 3 but it is still higher than the economical goal (17 yen/Nm 3 ). The major reason of the high cost comes from the large size of hydrogen separation reformers because of the limit of hydrogen separation efficiency of palladium membrane. A new highly efficient hydrogen separation membrane is needed to reduce the cost of hydrogen production using membrane reformers. There is possibility of multi-tube failure in the membrane reformers. In future study, a design of measures against tube failure and elemental experiments of reaction between sodium and reforming gas will be needed. (authors)

  16. Design and safety aspect of lead and lead-bismuth cooled long-life small safe fast reactors for various core configurations

    International Nuclear Information System (INIS)

    Zaki, S.; Sekimoto, Hiroshi

    1995-01-01

    Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ∼2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25-40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient. (author)

  17. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  18. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  19. Analyses for experiment on sodium-water reaction temperature by the CHAMPAGNE code

    International Nuclear Information System (INIS)

    Yoshioka, Naoki; Kishida, Masako; Yamada, Yumi

    2000-03-01

    In this work, analyses on sodium-water reaction temperature in the new SWAT-1(SWAT-1R) test were completed by the CHAMPAGNE code in order to understand void and velocity distribution in sodium system, which was difficult to be measured in experiments. The application method of the RELAP5/Mod2 code was investigated to LMFBR steam generator (SG) blow down analysis, too. The following results were obtained. (1) Analyses on sodium-water reaction temperature in the SWAT-1R test. 1) Analyses were carried out for the SWAT-1R test under the condition water leak rate 600 g/s by treating the pressure loss coefficient, the interface friction coefficient and the coefficient related to reaction rate as parameters. The effect and mechanism of each parameter on the shape of reaction zone were well understood by these analyses. 2) The void and velocity distribution in sodium system were estimated by use of the most suitable parameters. These analytical results are expected to be useful for planning of the SWAT-1R test and evaluation of test result. (2) Investigation of the RELAP5/Mod2 code. 1) The items to be improved in the RELAP5/Mod2 code were clarified to apply this code to the FBR SG blow down analysis. 2) One of these items was an addition of the shell-side (sodium-side) model. A sodium-side model was designed and added to the RELAP5/Mod2 code. Test calculations were carried out by this improved code and the basic function of this code was confirmed. (author)

  20. Enhancing MA transmutation by irradiation of (MA, Zr)Hx in FBR blanket region - 5383

    International Nuclear Information System (INIS)

    Konashi, K.; Ikeda, K.; Itoh, K.; Hirai, M.; Koyama, T.; Kurosaki, K.

    2015-01-01

    Minor actinide (MA) hydride is proposed as transmutation target in sodium-cooled mixed oxide fuelled fast reactor. Preliminarily calculations have been done to check the transmutation efficiency of MA hydride targets. Three different types of MA target, MA-Zr alloy, (MA, Zr)O 2 and (MA, Zr)H x , have been compared on MA transmutation rate. The targets are assumed to be loaded around an active core in a 280 MWe sodium-cooled reactor; 54 MA target assemblies are respectively arranged in a row in the radial blanket zone. They are supposed to be irradiated for one year and then be cooled for 60 days. The transmuted mass has been evaluated by three-dimensional diffusion calculation to be 25, 15, 61 kg/EFPY for the alloy, the oxide and the hydride respectively, where production of MA in the active core is taken into account. The transmutation mass by (MA, Zr)H x is much larger than those by the other types of targets, while the core characteristics remain sound by locating MA targets outside of the active core. On top of that, two kinds of (MA, Zr)O 2 targets which are combined with ZrH x (x=1.7) pins have been calculated. Major Research/Development items are selected to establish the MA hydride transmutation method by reviewing technologies applicable to the transmutation system. The practical use of the MA hydride transmutation method is not far ahead technically, since this method can be developed by the extension of existing technologies. (authors)

  1. C-Scan Performance Test of Under-Sodium ultrasonic Waveguide Sensor in Sodium

    International Nuclear Information System (INIS)

    Joo, Young Sang; Bae, Jin Ho; Kim, Jong Bum

    2011-01-01

    Reactor core and in-vessel structures of a sodium-cooled fast (SFR) are submerged in opaque liquid sodium in the reactor vessel. The ultrasonic inspection techniques should be applied for observing the in-vessel structures under hot liquid sodium. Ultrasonic sensors such as immersion sensors and rod-type waveguide sensors have developed in order to apply under-sodium viewing of the in-vessel structures of SFR. Recently the novel plate-type ultrasonic waveguide sensor has been developed for the versatile application of under-sodium viewing in SFR. In previous studies, the ultrasonic waveguide sensor module was designed and manufactured, and the feasibility study of the ultrasonic waveguide sensor was performed. To improve the performance of the ultrasonic waveguide sensor in the under-sodium application, a new concept of ultrasonic waveguide sensors with a Be coated SS304 plate is suggested for the effective generation of a leaky wave in liquid sodium and the non-dispersive propagation of A 0 -mode Lamb wave in an ultrasonic waveguide sensor. In this study, the C-scan performance of the under-sodium ultrasonic waveguide sensor in sodium has been investigated by the experimental test in sodium. The under-sodium ultrasonic waveguide sensor and the sodium test facility with a glove box system and a sodium tank are designed and manufactured to carry out the performance test of under-sodium ultrasonic waveguide sensor in sodium environment condition

  2. W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance

    International Nuclear Information System (INIS)

    Meyers, S.C.; Henderson, J.M.

    1980-05-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment

  3. Sodium-immersed self-cooled electromagnetic pump design and development of a large-scale coil for high temperature

    International Nuclear Information System (INIS)

    Oto, Akihiro; Naohara, Nobuyuki; Ishida, Masayoshi; Katsuki, Kenji; Kumazawa, Ryouji

    1995-01-01

    A sodium-immersed, self-cooled electromagnetic (EM) pump was recently studied as a prospective innovative technology to simplify a fast breeder reactor plant system. The EM pump for a primary pump, a pump type, was designed, and the structural concept and the system performance were clarified. For the flow control method, a constant voltage/frequency method was preferable from the point of view of pump performance and efficiency. The insulation life was tested on a large-scale coil at high temperature as part of the development of a large-capacity EM pump. Mechanical and electrical damage were not observed, and the insulation performance was quite good. The insulation system could also be applied to large-scale coils

  4. MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

    International Nuclear Information System (INIS)

    Sugino, K.; Iwai, T.

    2006-01-01

    MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

  5. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  6. Method for controlling FBR type reactor

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

    1991-01-01

    The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

  7. A Review of PSA Technology Applications according to the Development of Sodium-cooled Fast Reactors in the World

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Lee, Yong Bum; Jung, Hae Yong; Kim, Sang Ji; Hahn, Do Hee; Yang, Joon Eon

    2008-12-01

    The international nuclear societies request to perform Probabilistic Safety Assessment (PSA) according to the development of Gen IV Sodium-cooled Fast Reactors (SFR). One of the major tasks of the PSA is to identify various sequences of events which could lead to the release of radioactivity. However, due to the limited operating and SFR PSA experiences, it will be difficult to derive and to quantify core damage frequency for SFR under development in Korea, so called KALIMER. Hence, in this report, the foreign PSA results, such as USA and Japan, are analyzed based on the obtained documents. Finally an approach on how to perform PSA for KALIMER is suggested

  8. Specialists meeting on sodium removal and decontamination. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-08-01

    This report covers experiences on sodium removal techniques developed or gained in a number of countries running sodium cooled reactors. This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Cleaning of sodium wetted components or fuel assemblies was achieved by applying different techniques including vacuum distillation, using different alcohols or evaporation processes.

  9. Specialists meeting on sodium removal and decontamination. Summary report

    International Nuclear Information System (INIS)

    1978-08-01

    This report covers experiences on sodium removal techniques developed or gained in a number of countries running sodium cooled reactors. This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Cleaning of sodium wetted components or fuel assemblies was achieved by applying different techniques including vacuum distillation, using different alcohols or evaporation processes

  10. Performance of the diffusion barrier in the metallic fuel in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Ryu, Ho Jin; Yang, Seong Woo; Lee, Byoung Oon; Oh, Seok Jin; Lee, Chan Bock; Hahn, Dohee

    2009-01-01

    The objectives in this study are to propose several kinds of barrier materials and to evaluate their performance to prevent a fuel-clad interaction situation between the metallic fuel and the clad material in the Sodium-cooled Fast Reactor (SFR). Metallic foil made from refractory element, electrodeposition of the Cr on the clad surface, and the vapor deposition of the Zr were used as the barrier layers. The diffusion couple test was performed at the temperature of 800degC for 25 hour. The results showed that considerable amount of reaction occurred at the specimen without barrier, whereas excellent performance was observed in that neither reaction nor inter-diffusion occurred in the case of metallic foil made of Cr or V. Electrodeposition was revealed to be excellent provided that optimum deposition condition can be found. Similar to the electro-deposition result, excellent performance observed in the case of vapor deposition condition. (author)

  11. Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting

    International Nuclear Information System (INIS)

    Jong Hwan Kim; Hoon Song; Ki Hwan Kim; Chan Bock Lee

    2014-01-01

    Metal fuel slugs of U-Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U-10 wt% Zr and U-10 wt% Zr-5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions. (author)

  12. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  13. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  14. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  15. Friction and wear in sodium

    International Nuclear Information System (INIS)

    Hoffman, N.J.; Droher, J.J.

    1973-01-01

    In the design of a safe and reliable sodium-cooled reactor one of the more important problem areas is that of friction and wear of components immersed in liquid sodium or exposed to sodium vapor. Sodium coolant at elevated temperatures may severely affect most oxide-bearing surface layers which provide corrosion resistance and, to some extent, lubrication and surface hardness. Consequently, accelerated deterioration may be experienced on engaged-motion contact surfaces, which could result in unexpected reactor shutdown from component malfunction or failure due to galling and seizure. An overall view of the friction and wear phenomena encountered during oscillatory rubbing of surfaces in high-temperature, liquid-sodium environments is presented. Specific data generated at the Liquid Metal Engineering Center (LMEC) on this subject is also presented. (U.S.)

  16. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  17. Investigation of 144Pr measurements for determination of residual Pu in FBR leached hulls

    International Nuclear Information System (INIS)

    Yates, A.; Bremner, W.B.

    1983-12-01

    The measurement of plutonium in leached hulls arising from FBR fuel reprocessing is required for plant control, accountancy and safeguards. At DNPDE these measurements are carried out on the batch of hulls prior to bulking into 200 l drums for retrievable storage and ultimately further treatment for plutonium recovery. The experience to date has related to the use of neutron interrogation using sealed tube neutron generators as the irradiation source. The supply of replacement sealed tubes has become difficult. It was therefore decided to consider, amongst other techniques, the measurement of 144 Pr gamma emission as a possible alternative. The technique has been extensively used for thermal reactor fuels on a routine basis. There has also been a limited amount of work reported from Cap La Hague on applying the technique to FBR fuels as part of an experimental programme. This paper therefore describes the work done at DNPDE which evaluated the technique for use on a batch of hulls arising from a PFR fuel reprocessing campaign. (author)

  18. Theoretical study on flow-induced vibration of a cylindrical weir due to fluid discharge

    International Nuclear Information System (INIS)

    Fujita, Katsuhisa; Ito, Tomohiro; Hirota, Kazuo; Kodama, Tetsuhiko

    1994-01-01

    In a FBR, the inside of the reactor vessel is cooled by liquid sodium. Liquid sodium is supplied to the upper plenum from its bottom and discharges over the top of the cylindrical weir down to the lower plenum. The weir is so thin in order to decrease the thermal stress on it that the fluid--structure interaction becomes predominant. A fluidelastic vibration of the weir due to fluid discharge was discovered in a French FBR. In this study, a theoretical model was developed on the ''fluid--elastic mode'' instability of a cylindrical weir due to fluid discharge from the upper plenum to the lower plenum. In the analysis, the fluctuation of both the discharge flow rate over a weir due to the vibration of the cylindrical shell and the pressure in the lower plenum due to fluid discharge were formulated. Instability criteria was derived from the added damping ratio due to fluid discharge using modal analysis. The natural modes and modal mass of the weir were obtained by the analysis using the FEM code taking the fluid - structure interaction into consideration. The theoretical instability range in terms of the fall height and the flow rate is compared with the experimental results. The theoretical values showed a good agreement with the experimental ones

  19. Local transport of vertically- and horizontally-emitted sodium oxide aerosols

    International Nuclear Information System (INIS)

    Fields, D.E.; Miller, C.W.; Cooper, A.C.

    1986-01-01

    Liquid-metal cooled breeder reactors are expected to use large quantities of sodium or sodium-potassium alloy, and evaluation of the possible consequences of a liquid-metal fire, henceforth referred to as a sodium fire, is an important consideration. Of particular interest is the sodium aerosol concentration at the air intake ports that are used for reactor cooling, and which might suffer restricted flow under high aerosol concentrations. We have devised and applied a methodology for estimating the concentration of aerosols released vertically and horizontally from building surfaces and monitored at other building surface points. We have used this methodology to make calculations that indicate the time-development of aerosol build-up, and the maximum aerosol concentrations, at air intake ports. Building wake effects, momentum-driven plume rise, and density-driven plume rise are considered

  20. Local transport of vertically and horizontally emitted sodium oxide aerosols

    International Nuclear Information System (INIS)

    Fields, D.E.; Miller, C.W.; Cooper, A.C.

    1986-01-01

    Liquid-metal-cooled breeder reactors are expected to use large quantities of sodium or sodium-potassium alloy, and evaluation of the possible consequences of a liquid-metal fire, henceforth referred to as a sodium fire, is an important consideration. Of particular interest is the sodium aerosol concentration at the air intake ports that are used for reactor cooling, and which might suffer restricted flow under high aerosol concentrations. The authors have devised and applied a methodology for estimating the concentration of aerosols released vertically and horizontally from building surfaces and monitored at other building surface points. This methodology has been used to make calculations that indicate the time development of aerosol buildup, and the maximum aerosol concentration, at air intake ports. Building wake effects, momentum-driven plume rise, and density-driven plume rise are considered

  1. JAEA FBR Plant Engineering Center annual report 2011

    International Nuclear Information System (INIS)

    2012-11-01

    The FBR Plant Engineering Center was established on April 1, 2009 located in a research building, of which care is taken by the International Nuclear Information Training Center, Tsuruga Head Office, at Shiraki in Tsuruga. The mission of the center is to perform R and D (research and development) works both for analysis of operational experiences at the prototype fast breeder reactor “Monju” and for technology development concerning design and operation of “Monju”. Moreover it is also required to apply the results to next generation fast breeder reactors, which is an important role of Advanced Nuclear System Research and Development Directorate. And in these R and D activities, it is expected to conduct the works in cooperation with domestic or foreign research organizations or universities by a joint-study or a collaborative-work manner. The R and D activities have been carried out specifically on the “demonstration of the reliability as a power generation plant” and “establishment of sodium handling technology”, which are originally intended missions of “Monju”. And the other R and Ds have been promoted both for the plant engineering, such as plant maintenance, to effectively use an existing reactor in order to apply the R and D results to a future demonstration reactor, and for the irradiation test study, such as advanced fuel irradiation, to use “Monju” as an irradiation test bed. In order to perform these R and D activities, five R and D groups have been set up in the center. They are operation-and-maintenance engineering, sodium engineering, reactor-core-and-fuel engineering, plant engineering, and safety engineering groups. However, the Japanese atomic energy policy is being reviewed after the accident of the Fukushima Daiichi nuclear power station caused by a tsunami generated by the Tohoku-district-off-the-Pacific-Ocean Earthquake on March 11, 2011, and all the R and D activities using “Monju” have been suspended since late 2011

  2. Inverted Steam Generators for Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Matal, Oldřich; Šimo, Tomáš; Matal, Oldřich Jr.

    2013-01-01

    Conclusions: Two inverted steam generators of the Czech industry provenience have still been in successful operation with no water into sodium leaks at BOR 60 (RIAR Dimitrovgrad, Russian Federation). Micromodular inverted steam generator (MMISG) since 1981 and modular inverted steam generator (MISG) since 1991. In the framework of the CP ESFR project predesign studies of 100 MW (thermal) ISG modules were performed with the consideration of MMISG and MISG design, operational and safety benefits and experience. Development of material and technology for sodium heated steam generators components reflecting contemporary domestic industrial conditions in the Czech Republic was restarted in the years 2003 to 2004 and supported in the years 2008 to 2011 by the European CP ESFR project and by the Ministry of Industry and Trade of the Czech Republic

  3. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  4. Non-aqueous removal of sodium from reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Welch, F H; Steele, O P [Rockwell International, Atomics International Division, Canoga Park (United States)

    1978-08-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component.

  5. Non-aqueous removal of sodium from reactor components

    International Nuclear Information System (INIS)

    Welch, F.H.; Steele, O.P.

    1978-01-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component

  6. Technical meeting on decommissioning of fast reactors after sodium draining. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    The objective of the technical meeting was to provide a forum for in-depth scientific and technical exchange on topics related to the decommissioning experience with fast reactors, in particular with regard to the decommissioning of components after sodium draining. Accordingly, the scope of the meeting covers the review and analyses of the experience gained from the decommissioning of both active sodium loops and sodium cooled fast reactors (e.g., KNK II, Superphenix, RAPSODIE, EBR-II, FERMI, BN-350, BR-10). It is expected that the outcome of the meeting will contribute to the Agency initiative to preserve fast reactor data and knowledge. The main focus of the technical meeting was given on the decommissioning of both active loop and reactor components (e.g., the primary vessel of a sodium-cooled reactor) that have been drained of sodium, but that still conserve some residual amounts of sodium (e.g., films covering the entire surface of the component, or particular sodium heels that cannot be drained)

  7. Review of the sodium fire experiments including sodium-concrete-reactions and summary of the results

    International Nuclear Information System (INIS)

    Cherdron, W.

    1996-01-01

    In the technical and design concept of containment systems of sodium cooled breeder reactors it has to be considered, that leakages in sodium pipes lead to sodium fires. The temperature and pressure rise caused by sodium fires makes it indispensable to analyse these accidents to be able to assess the safety of the whole system. Generally sodium leakages may lead to three different types of fires with different consequences. The main influences are the geometry of the leakage, shape, size, location, and the sodium conditions, such as temperature, flow rate and velocity. It must be also considered the reaction of sodium with surfaces like concrete. The paper gives an overview over all the sodium fire experiments performed in the FAUNA-facility (220 m 3 ) of the Forschungszentrum Karlsruhe in the years 1979 to 1993. The experimental program started with the investigation of pool fires on burning areas between 2 and 12 m 2 with up to 500 kg of Sodium. The experiments had been continued with 3 combined fires and 40 experiments on spray fires. 7 experiments on sodium-concrete reactions completed the program. (author)

  8. Experimental study associated to irradiation of FBR structural material, (4)

    International Nuclear Information System (INIS)

    1976-01-01

    The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

  9. An experimental study on impingement wastage of Mod 9Cr 1Mo steel due to sodium water reaction

    Energy Technology Data Exchange (ETDEWEB)

    Kishore, S., E-mail: skishore@igcar.gov.in [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Ashok Kumar, A.; Chandramouli, S.; Nashine, B.K.; Rajan, K.K.; Kalyanasundaram, P.; Chetal, S.C. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer Sodium heated steam generators are crucial components of fast breeder reactors. Black-Right-Pointing-Pointer A leak in steam generator tube will cause sodium water reaction that damages the tubes. Black-Right-Pointing-Pointer Experimental study was conducted to quantify the extent of damage on Mod 9Cr 1Mo tube due to a water leak. - Abstract: Sodium heated steam generator (SG) is a crucial component in the heat transport system of a fast breeder reactor (FBR). In case, one of its water/steam carrying tubes becomes defective, water/steam leaks into sodium, flowing in the shell side, causing sodium-water reaction, which is highly exothermic and producing corrosive NaOH. The reaction jet originating from a leaking tube may impinge on its adjacent tube, resulting in damage of the tube. Impingement wastage refers to this kind of damage, occurring to a tube of sodium heated SG, owing to a small water/steam leak from a neighboring tube. Extensive research works have been conducted all over the world to study various aspects of this phenomenon. Experimental studies were carried out in Indira Gandhi Centre for Atomic Research (IGCAR) to understand the effect of impingement wastage on Mod 9Cr 1Mo, which is the tube material of prototype fast breeder reactor (PFBR) SG. This paper brings out the data and experience gained through the experiments.

  10. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  11. Developing maintenance technologies for FBR's heat exchanger units by advanced laser processing

    International Nuclear Information System (INIS)

    Nishimura, Akihiko; Shimada, Yukihiro

    2011-01-01

    Laser processing technologies were developed for the purpose of maintenance of FBR's heat exchanger units. Ultrashort laser processing fabricated fiber Bragg grating sensor for seismic monitoring. Fiber laser welding with a newly developed robot system repair cracks on inner wall of heat exchanger tubes. Safety operation of the heat exchanger units will be improved by the advanced laser processing technologies. These technologies are expected to be applied to the maintenance for the next generation FBRs. (author)

  12. Development of wire wrapping technology for FBR fuel pin

    International Nuclear Information System (INIS)

    Nogami, Tetsuya; Seki, Nobuo; Sawayama, Takeo; Ishibashi, Takashi

    1991-01-01

    For the FBR fuel assembly, the spacer wire is adopted to maintain the space between fuel pins. The developments have been carried out to achieve automatically wire wrapping with high precision. Based on the fundamental technology developed through the mock-up test operation, Joyo 'MK-I', fuel pin fabrication was started using partially mechanized wire wrapping machine in 1973. In 1978, an automated wire wrapping machine for Joyo 'MK-II' was developed by the adoption of some improvements for the wire inserting system to end plug hole and the precision of wire pitch. On the bases of these experiences, fully automated wire wrapping machine for 'Monju' fuel pin was installed at Plutonium Fuel Production Facility (PFPF) in 1987. (author)

  13. Future work in the DeBeNeLux research centres on the sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Goedkoop, J.A.

    1976-01-01

    The general objectives as they now apply over the world in the further development of the sodium cooled fast reactor are to realize a reactor and the associated fuel cycle, that will ensure a good fuel utilization; secondly, as long as we live in a more or less free market economy, such a system will only be acceptable if it is competitive, which means that the difference in investment cost between the fast reactor and the presently used light water reactors has to be brought down; thirdly, to justify the investment the system should work reliably; finally the developments in reactor design should not be at the expense of reactor safety. The pursuit of these objectives during the coming years will require the DeBeNeLuX laboratories to do work in a number of fields. (Auth.)

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

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  16. Application of the PSA method to decay heat removal systems in a large scale FBR design

    International Nuclear Information System (INIS)

    Kotake, S.; Satoh, K.; Matsumoto, H.; Sugawara, M.; Sakata, K.; Okabe, A.

    1993-01-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10 -7 /d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  17. Application of the PSA method to decay heat removal systems in a large scale FBR design

    Energy Technology Data Exchange (ETDEWEB)

    Kotake, S; Satoh, K [Japan Atomic Power Company, Otemachi, Chiyoda-ku, Tokyo (Japan); Matsumoto, H; Sugawara, M [Toshiba Corporation (Japan); Sakata, K [Mitsubishi Atomic Power Industries Inc. (Japan); Okabe, A [Hitachi Engineering Co., Ltd. (Japan)

    1993-02-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10{sup -7}/d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  18. FBR Italian position. The PEC reactor and its control rods system

    International Nuclear Information System (INIS)

    Artioli, C.

    1984-01-01

    The greatest effort in Italy concerning Fast Breeder Reactors is concentrated on the PEC project. This project represents in fact the sole Italian work on the FBR option. Up to now the state of progress of the PEC project is as follows: - Design work on reactor components and related systems: 80% - Civil engineering work: 45% - Components construction: 18%. It has recently been stated that construction of the plant must be finished by 1987. The progressive design of the PEC control system and the supporting research and development programmes are described in this paper

  19. Feasibility Study on Ultrasonic Waveguide Sensor for Under-Sodium Visualization of Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young-Sang; Park, Chang-Gyu; Lee, Jae-Han; Lim, Sa-Hoe

    2008-01-15

    The reactor core and internal structures of a sodium-cooled fast reactor (SFR) can not be visually examined due to the opaque sodium. The examination of the internal structures is possible by using ultrasonics to penetrate the sodium. The under-sodium viewing technique using an ultrasonic wave should be applied for the in-service inspection of the reactor internals. Immersion sensors and waveguide sensors have been utilized for the under-sodium viewing application. The immersion sensor has a precise imaging capability, but may have high temperature restrictions and an uncertain life. The waveguide sensor can operate in a hostile environment, such as liquid metal at a high temperature in the presence of high radiation. The waveguide sensor has the advantages of simplicity and reliability, but limits in its movement. A new plate-type waveguide sensor has been developed to overcome the limitations of previous waveguide sensors. And a novel ultrasonic technique has been suggested. The technique is capable of steering a radiation beam of a waveguide sensor without a mechanical movement of the waveguide sensor. The control of the radiation beam angle can be achieved by a frequency tuning method of the excitation pulse in the dispersive low frequency range of the A{sub 0} Lamb wave. A waveguide sensor assembly has been designed for the actual application of undersodium visual inspection in sodium-cooled fast reactor. The main purpose of this study is achievement of feasibility of ultrasonic waveguide sensor technology to the application of undersodium viewing. Under-water C-scan imaging test was carried out by using 10 m long waveguide sensor assembly. It was confirmed that the test target could be clearly visualized and the resolution of C-scan image could be less than 2 mm.

  20. Feasibility Study on Ultrasonic Waveguide Sensor for Under-Sodium Visualization of Sodium Fast Reactor

    International Nuclear Information System (INIS)

    Joo, Young-Sang; Park, Chang-Gyu; Lee, Jae-Han; Lim, Sa-Hoe

    2008-01-01

    The reactor core and internal structures of a sodium-cooled fast reactor (SFR) can not be visually examined due to the opaque sodium. The examination of the internal structures is possible by using ultrasonics to penetrate the sodium. The under-sodium viewing technique using an ultrasonic wave should be applied for the in-service inspection of the reactor internals. Immersion sensors and waveguide sensors have been utilized for the under-sodium viewing application. The immersion sensor has a precise imaging capability, but may have high temperature restrictions and an uncertain life. The waveguide sensor can operate in a hostile environment, such as liquid metal at a high temperature in the presence of high radiation. The waveguide sensor has the advantages of simplicity and reliability, but limits in its movement. A new plate-type waveguide sensor has been developed to overcome the limitations of previous waveguide sensors. And a novel ultrasonic technique has been suggested. The technique is capable of steering a radiation beam of a waveguide sensor without a mechanical movement of the waveguide sensor. The control of the radiation beam angle can be achieved by a frequency tuning method of the excitation pulse in the dispersive low frequency range of the A 0 Lamb wave. A waveguide sensor assembly has been designed for the actual application of undersodium visual inspection in sodium-cooled fast reactor. The main purpose of this study is achievement of feasibility of ultrasonic waveguide sensor technology to the application of undersodium viewing. Under-water C-scan imaging test was carried out by using 10 m long waveguide sensor assembly. It was confirmed that the test target could be clearly visualized and the resolution of C-scan image could be less than 2 mm

  1. Inner wall attack and its inhibition method for FBR fuel pin cladding at high burnup

    International Nuclear Information System (INIS)

    Xu Yongli; Long Bin; Li Jingang; Wan Jiaying

    1998-01-01

    The inner wall attack of the modified 316-Ti S.S. cladding tubes manufactured in China used FBR at 10at.% burnup was investigated by means of the out of pile simulation tests. The inner surface morphologies of the cladding tubes attached by fission products Cs, Te, I and Se at 700 deg. C under lower and high oxygen potentials were observed respectively, and the depth of attack was also measured. The burst strength, maximum circum expansion and the appearances of fracture were measured and observed respectively for the cladding tubes attacked by fission products. Based on the mechanism of FBR fuel cladding chemical interaction (FCCI), Cr, Zr and Nb were used as the oxygen absorbers respectively, in order to inhibit the inner wall attack of the cladding tubes. The corrosion morphologies and depth, the penetration depth of the fission products in the inner surface of the cladding tubes were detected. The inhibition effectiveness of the oxygen absorbers for the inner wall attack of the cladding tubes was evaluated. (author)

  2. Development of simulation code for FBR spent fuel dissolution with rotary drum type continuous dissolver

    International Nuclear Information System (INIS)

    Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi

    2011-01-01

    Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO 2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (∼500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

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

  5. Development of a sodium ionization detector for sodium-to-gas leaks

    International Nuclear Information System (INIS)

    Swaminathan, K.; Elumalai, G.

    1984-01-01

    A sensitive sodium-to-gas leak detector has been indigenously developed for use in liquid metal cooled fast breeder reactor. The detector relies on the relative ease with which sodium vapour or its aerosols including its oxides and hydroxides can be thermally ionized compared with other possible constituents such as nitrogen, oxygen, water vapour etc. in a carrier gas and is therefore called sodium ionization detector (SID). The ionization current is a measure of sodium concentration in the carrier gas sampled through the detector. Different sensor designs using platinum and rhodium as filament materials in varying sizes were constructed and their responses to different sodium aerosol concentrations in the carrier gas were investigated. Nitrogen was used as the carrier gas. Both the background current and speed of response were found to depend on the diameter of the filament. There was also a particular collector voltage which yielded maximum sensitivity of the detector. The sensor was therefore optimised considering influence of above factors and a detector has been built which demonstrates a sensitivity better than 0.3 nanogram of sodium per cubic centimetre of carrier gas for a signal to background ratio of 1:1. Its usefulness in detecting sodium fires in experimental area was also demonstrated. Currently efforts are under way to improve the life time of the filament used in the above detector. (author)

  6. Safety approach and R and D program for future french sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Beils, Stephane; Carluec, Bernard; Devictor, Nicolas; Fiorini, Gian Luigi; Sauvage, Jean Francois

    2011-01-01

    This paper presents briefly the safety approach as well as the R and D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R and D. B) Strategy and roadmap in support of the R and D for future SFRs. This section describes the R and D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype. (author)

  7. Large electro-magnetic pump design for application in the ASTRID sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Laffont, Guy; Rey, Frédéric; Aizawa, Rie; Suziki, Tetsu

    2013-01-01

    Conclusion: • Use of a LEMP motivated by several advantages in terms of the reactor design, operation and maintenance. • Collaboration agreement between the CEA and TOSHIBA Corporation came into force in April 2012 to carry out a joint work program on the ASTRID EMP design and development. • Preliminary LEMP calculations carried out by the CEA and TOSHIBA are in good agreement and provide a good confidence in the feasibility of the annular LEMP for the ASTRID intermediate sodium loop. • Theoretical and experimental investigations are currently underway at the CEA with the aim to improve the numerical tools. • In parallel, the ASTRID EMP conceptual design studies are ongoing at TOSHIBA (thermal and thermo-mechanical analyses to demonstrate the LEMP self-cooling, structural analysis of the casing, the supporting legs and the mechanical interfaces, definition of the power supply unit, instrumentation and remote control procedure). • This program is aiming at consolidating the ASTRID EMP conceptual design report and to support the design option choice for the ASTRID basic design

  8. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  9. Fertile assembly for a fast neutron nuclear reactor cooled by liquid sodium, with regulation of the cooling rate

    International Nuclear Information System (INIS)

    Pradal, L.; Berte, M.; Chiarelli, C.

    1985-01-01

    The assembly has a casing in which are arranged the fertile elements, the liquid sodium flowing through the casing along these elements. It includes several apertured diaphragms transverse to the rods to regulate the liquid sodium flow rate. At least one diaphragm, in its central part around its aperture, of a material soluble in liquid sodium, such as copper. The invention applies, more particularly, to fast neutron nuclear reactor having a heterogeneous core. The coolant flow can increase with time to match the increased power generated by the fertile assembly along its life [fr

  10. JSFR design progress related to development of safety design criteria for Generation IV sodium-cooled fast reactors. (1) Overview

    International Nuclear Information System (INIS)

    Kamide, Hideki; Ando, Masato; Ito, Takaya

    2015-01-01

    JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R and D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a looptype next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500 MWe for the commercial version and 750 MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR. (author)

  11. Study on enhancement of heat transfer of reactor vessel auxiliary cooling system of fast breeder reactor

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Kinoshita, Izumi; Ueda, Nobuyuki; Furuya, Masahiro

    1996-01-01

    A reactor vessel auxiliary cooling system (RVACS), which is one of the decay heat removal systems of the fast breeder reactor (FBR), has passive safety as well as high reliability. However, the heat removal capability is relatively small, because its heat exchange is dependent on the natural convection of the air. The objectives of this report are to propose a heat transfer medium to enhance the heat transfer and to confirm the heat transfer performance of this system by experimental and analytical studies. From these studies, the following main results were obtained. (1) A porous plate with 5 mm thickness, 5 mm pore diameter, 92% porosity, was found to have the highest enhancement of heat transfer. (2) The heat transfer enhancement was demonstrated by large scale heat transfer experiments. Also, the heat transfer correlations, which can be used in the plant transient analyses, were derived from the experimental results. (3) Analysing the transient conditions of conventional pool-type FBR by means of the system analysis code, the applicable range of this system was assumed from the capability of the RVACS with porous plates. As a result, this type of RVACS was found to be applicable to conventional pool-type FBRs with capacity of about 500 MWe or less. (author)

  12. Fuel assembly cooling experience at the FFTF/IEM cell

    International Nuclear Information System (INIS)

    McGuinness, P.W.

    1985-01-01

    In the Fast Flux Test Facility (FFTF), sodium wetted irradiated fuel assemblies are discharged to the Interim Examination and Maintenance (IEM) Cell for disassembly and post-irradiation examination in an inert argon atmosphere. While in the IEM Cell, fuel assemblies are cooled by the IEM Cell Subassembly Cooling System. This paper describes the cooling system design, performance, and lessons learned, including a discussion of two overtemperature incidents. 2 refs., 6 figs

  13. Chinese FBR program and its first FR conceptual design

    International Nuclear Information System (INIS)

    Xu Mi

    1991-01-01

    As the preliminary study results, The long term strategy of FBR development in China would be divided into three steps: Experimental Fast Reactor, Modulized Fast Breeder Reactor and Large Fast Breeder Reactor. The design requirements, main technical selections and design boundary conditions for the first step, experimental fast reactor, which is named FFR (First Fast Reactor) in China have been given in the paper. The conceptual design of FFR, based on core neutronics, core subassemblies thermo-hydraulics, reactor block and mechanics, fuel handling system, main heat transfer system and steam-electricity generation system has been completed and also presented in this paper. Finally, the R and D program of FFR which is carring on and will be done is briefly introduced. (author)

  14. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  15. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  16. Emergency cooling apparatus for reactor

    International Nuclear Information System (INIS)

    Sakaguchi, S.

    1975-01-01

    A nuclear reactor is described which has the core surrounded by coolant and an inert cover gas all sealed within a container, an emergency cooling apparatus employing a detector that will detect cover gas or coolant, particularly liquid sodium, leaking from the container of the reactor, to release a heat exchange material that is inert to the coolant, which heat exchange material is cooled during operation of the reactor. The heat exchange material may be liquid niitrogen or a combination of spheres and liquid nitrogen, for example, and is introduced so as to contact the coolant that has leaked from the container quickly so as to rapidly cool the coolant to prevent or extinguish combustion. (Official Gazette)

  17. Investigation of sodium area conflagrations and testing of a protective system

    Energy Technology Data Exchange (ETDEWEB)

    Huber, F; Menzenhauer, P; Peppler, W [Kernforschungszentrum Karlsruhe (F.R. Germany). Inst. fuer Reaktorentwicklung

    1975-12-01

    During research and development work on the SNR-300 sodium-cooled fast reactor the consequences and confinement of sodium fires occurring in enclosures were studied. The behavior of liquid sodium during fires and the behavior of an inherently ready-for-operation protective system are described. Theoretical considerations on the behavior of burning liquid sodium are compared with experimental results. A protective system for large facilities is presented and the use of extinguishing powders is reviewed.

  18. Preliminary conceptual design of the secondary sodium circuit-eliminated JSFR (Japan Sodium Fast Reactor) adopting a supercritical CO2 turbine system (1). Sodium/CO2 heat exchanger

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Sakamoto, Yoshihiko; Kotake, Shoji

    2014-09-01

    Research and development of the supercritical CO 2 (S-CO 2 ) cycle turbine system is underway in various countries for further improvement of the safety and economy of sodium-cooled fast reactors. The Component Design and Balance-Of-Plant (CD and BOP) of the Generation IV International Nuclear Forum (Gen-IV) has addressed this study, and their analytical and experimental results have been discussed between the relevant countries. JAEA, who is a member of the CD and BOP, has performed a design study of an S-CO 2 gas turbine system applied to the Japan Sodium-cooled Fast Reactor (JSFR). In this study, the S-CO 2 cycle turbine system was directly connected to the primary sodium system of the JSFR to eliminate the secondary sodium circuit, aiming for further economical improvement. This is because there is no risk of sodium-water reaction in the S-CO 2 cycle turbine system of SFRs. The Na/CO 2 heat exchanger is one of the key components for the secondary sodium system eliminated SFR, and this report describes its structure and the safety in case of CO 2 leak. A Printed Circuit Heat Exchanger (PCHE), which has a greater heat transfer performance, is employed to the heat exchanger. Another advantage of the PCHE is to limit the area affected by a leak of CO 2 because of its partitioned flow path structure. A SiC/SiC ceramic composite material is used for the PCHE to prevent crack growth and to reduce thermal stress. The Na/CO 2 heat exchanger has been designed in such a way that a number of small heat transfer modules are combined in the vessel in consideration of manufacture and repair. The primary sodium pump is installed in the center of the heat exchanger vessel. CO 2 leak events in the heat exchanger have been also evaluated, and it revealed that no significant effect has arisen on the core or the primary sodium boundary. (author)

  19. Sodium-sodium intermediate heat exchangers design problems

    International Nuclear Information System (INIS)

    Chandramohan, R.

    1975-01-01

    This paper deals briefly with the calculation methods adapted, in working-out the stresses due to fluid pressures (normal as well as transient), weights, piping-reactions, vibration in the tube-bundle and also the thermal stresses during normal and transient conditions, for the mechanical design of intermediate heat-exchanger. The thermal stress evaluation of the tube-sheet is given particular emphasis. A brief outline of the design problems connected with the Na-Na exchangers of large size sodium cooled fast reactor plants is also given. (author)

  20. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

    Kuriakose, K.K.

    2013-01-01

    The Process Architecture: → Acquire: Solicitation; Voluntary submission; Mandatory requirements; Interview/Observation; → Quality Control: Review/Editing; Certification; Quality index; → Disseminate: Publish through the Technology architecture; Formal/Informal Meetings; COPs; → Utilize: Projects; Day-to-day activities; → Maintenance; → Retirement. Mission: To conduct a broad based multidisciplinary programme of scientific research and advanced engineering development, directed towards the establishment of the technology of Sodium Cooled Fast Breeder Reactors (FBR) and associated fuel cycle facilities in the Country. The mission includes the development and applications of new and improved materials, techniques, equipment and systems for FBRs, pursue basic research to achieve breakthroughs in Fast Reactor technology

  1. Corrosion and material transfer in a sodium loop

    International Nuclear Information System (INIS)

    Garcia, A.M.; Espigares, M.M.; Arroyo, J.; Borgstedt, H.U.; Kernforschungszentrum Karlsruhe G.m.b.H.

    1984-01-01

    The corrosion and material transfer behaviour of the martensitic steel X18 CrMoVNb 12 1 as a function of the temperature and the position is studied in the ML-1 sodium loop. Up to 600 C the material has the same good compatibility with liquid sodium as austenitic stainless steels, as well in the corrosion region of the loop as in the deposition zone in the cooled leg. The steel is not sensitive to carburization or decarburization under the conditions in the sodium rig. (author)

  2. Aiming at super long term application of nuclear energy. Scope and subjects on the water cooled breeder reactor, the 'reduced moderation water reactor'

    International Nuclear Information System (INIS)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2001-01-01

    In order to make possible on nuclear energy application for super long term, development of sodium cooling type fast breeder reactor (FBR) has been carried out before today. However, as it was found that its commercialization was technically and economically difficult beyond expectation, a number of nations withdrew from its development. And, as Japan has continued its development, scope of its actual application is not found yet. Now, a research and development on a water cooling type breeder reactor, the reduced moderation water reactor (RMWR)' using LWR technology has now been progressed under a center of JAERI. This RMWR is a reactor intending a jumping upgrade of conversion ratio by densely arranging fuel bars to shift neutron spectrum to faster region. The RMWR has a potential realizable on full-dress plutonium application at earlier timing through its high conversion ratio, high combustion degree, plutonium multi-recycling, and so on. And, it has also feasibility to solve uranium resource problem by realization of conversion ratio with more than 1.0, to contribute to super long term application of nuclear energy. Here was investigated on an effect of reactor core on RMWR, especially of its conversion ratio and plutonium loading on introduction effect as well as on how RMWR could be contributed to reduction of uranium resource consumption, by drawing some scenario on development of power generation reactor and fuel cycle in Japan under scope of super long term with more than 100 years in future. And, trial calculation on power generation cost of the RMWR was carried out to investigate some subjects at a viewpoint of upgrading on economy. (G.K.)

  3. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  4. Method of processing waste sodium

    International Nuclear Information System (INIS)

    Shimoyashiki, Shigehiro; Takahashi, Kazuo.

    1982-01-01

    Purpose: To enable safety store of waste sodium in the form of intermetallic compounds. Method: Waste sodium used in a reactor is mixed with molten metal under an inert gas atmosphere and resulted intermetallic compounds are stored in a closely sealed container to enable quasi-permanent safety store as inert compound. Used waste sodium particularly, waste sodium in the primary system containing radioactive substances is charged in a waste sodium melting tank having a heater on the side, the tank is evacuated by a vacuum pump and then sealed with gaseous argon supplied from a gaseous argon tank, and waste sodium is melted under heating. The temperature and the amount of the liquid are measured by a thermometer and a level meter respectively. While on the other hand, molten metal such as Sn, Pb and Zn having melting point above 300 0 C are charged in a metal melting tank and heated by a heater. The molten sodium and the molten metals are charged into a mixing tank and agitated to mix by an induction type agitator. Sodium vapors in the tank are collected by traps. The air in the tank is replaced with gaseous argon. The molten mixture is closely sealed in a drum can and cooled to solidify for safety storage. (Seki, T.)

  5. BWR noise spectra and application of noise analysis to FBR

    International Nuclear Information System (INIS)

    Nomura, T.

    1975-01-01

    Work related to noise analysis, in Tokyo Shibaura Electric Co. Ltd. (Toshiba) and Nippon Atomic Industry Group Co. Ltd. (NAIG) for the past several years is reviewed. After considering the Japan-United States Seminar on Reactor Noise Analysis in 1968, other subjects discussed were boiling water reactor noise analysis and work in relation to FBR. Parts of these are related to each other. For example, boiling detection and temperature fluctuations are problems pertinent to both fields. As the main problems in zero-power-reactor noise are now basically understood, only a brief description of the experiments involving the advanced two detector method is made. Focus is rather placed on the area of power plant noise. (author)

  6. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    International Nuclear Information System (INIS)

    Piet, Steven J.; Bays, Samuel E.; Pope, Michael A.; Youinou, Gilles J.

    2010-01-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  7. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    Energy Technology Data Exchange (ETDEWEB)

    Kuk, Seoung Woo, E-mail: swkuk@kaeri.re.kr [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Youn, Young-Sang [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Jong-Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Radiochemistry & Nuclear Nonproliferation, University of Science & Technology, Gajeong-ro 217, Yuseong-gu, Daejeon, 34113 (Korea, Republic of)

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  8. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

    A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

  9. Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

    2015-09-15

    Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

  10. Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

  11. Conceptual design for Japan Sodium-Cooled Fast Reactor. (4) Developmental study of steel plate reinforced concrete containment vessel for JSFR

    International Nuclear Information System (INIS)

    Hosoya, Takusaburo; Negishi, Kazuo; Satoh, Kenichiro; Somaki, Takahiro; Matsuo, Ippei; Shimizu, Katsusuke

    2009-01-01

    An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results. (author)

  12. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  13. Scenario for commercialization of fast breeder reactors

    International Nuclear Information System (INIS)

    Kumaoka, Yoshio; Sato, Morihiko

    1989-01-01

    To realize the commercialization of fast breeder reactors (FBRs), it is essential to reduce construction costs to the same level as those for the current light water reactors. For this target to be attained, a highly important factor is to reduce to the lowest-levels possible the quantities of materials and volume of the buildings required for the primary and secondary sodium loops of the FBR. In this direction, an innovative compact FBR plant concept which holds promise for commercialization has been developed by introducing the pooltype reactor concept with the shortest possible secondary sodium loops, realized by coupling electromagnetic pumps with the steam generators. In comparison with the French Super Phenix reactor, for example, the construction of this 1,300-MWe FBR plant could be achieved with half the material quantities and plant volume required by the former type. (author)

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

    International Nuclear Information System (INIS)

    Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

    2012-09-01

    The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

  15. Review of the Technical Status on the Debris Bed Cooling Model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

    2007-09-15

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris0.

  16. Review of the Technical Status on the Debris Bed Cooling Model

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

    2007-09-01

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris

  17. Temperature dependence on sodium-water chemical reaction

    International Nuclear Information System (INIS)

    Tamura, Kenta; Deguchi, Yoshihiro; Suzuki, Koichi; Takata, Takashi; Yamaguchi, Akira; Kikuchi, Shin; Ohshima, Hiroyuki

    2012-01-01

    In a sodium-cooled fast reactor (SFR), liquid sodium is used as a heat transfer fluid because of its excellent heat transport capability. On the other hand, it has strong chemical reactivity with water vapor. One of the design basis accidents of the SFR is the water leakage into the liquid sodium flow by a breach of heat transfer tubes. This process ends up damages on the heat transport equipment in the SFR. Therefore, the study on sodium-water chemical reactions is of paramount importance for security reasons. This study aims to clarify the sodium-water reaction mechanisms using laser diagnostics. A quasi one-dimensional flame model is also applied to a sodium-water counter-flow reaction field. Temperature, H 2 , H 2 O, OH, Na and Particulate matter were measured using laser induced fluorescence and CARS in the counter-flow reaction field. The temperature of the reaction field was also modified to reduce the condensation of Na in the reaction zone. (author)

  18. Material properties of oxide dispersion strengthened (ODS) ferritic steels for core materials of FBR. Mechanical strength properties of sodium exposed and Nickel diffused materials. Interim report

    International Nuclear Information System (INIS)

    Kato, Shoichi; Yoshida, Eiichi

    2004-02-01

    An oxide dispersion strengthened (ODS) ferritic steel have excellent resistance to swelling and superior creep strength, they are expected to be used as a long-life cladding material in future advanced fast reactor. In this study, sodium environmental effects on the ODS steel developed by JNC were clarified through tensile test after sodium exposure for maximum 10,000hrs and creep-rupture test in sodium at elevated temperature. The exposure to sodium was conducted using a sodium test loop constituted by austenitic steels. For the conditions of sodium exposure test, the sodium temperatures were 923 K and 973 K, the oxygen concentration in sodium was below 2ppm and sodium flow rate on the surface of specimen was less than 1x10 -4 m/s. Further the specimen with the nickel diffused was prepared, which is simulate to nickel diffusing through sodium from the surface of structural stainless steels. The main results obtained were as follows; (1) The results showed excellent sodium-resistance up to a high temperature of about 973 K in stagnant sodium conditions, and its considered that the effects of sodium environment of tensile properties were negligible. In case of stagnant sodium condition, creep-rupture strength in sodium was equal to the in argon gas, and no sodium environmental effect was observed. The same is true for the creep-rupture ductility. (2) The tensile properties of nickel diffused test specimens at high temperatures simulating microstructure change were equal to that of the thermal aging process specimens. These tensile tests suggest that sodium environmental effects can be ignored. However, the effect of nickel diffusion on creep strength are not clear at present and experimental investigation are being conducted. (3) The coefficient of nickel diffusion in the ODS steel can be estimated based on the results of nickel concentration measurement. This value is larger than that of the diffusion coefficient for typical α-Fe steel at temperature below 973 K

  19. Sodium purification in Rapsodie

    International Nuclear Information System (INIS)

    Giraud, B.

    1968-01-01

    This report is one of a series of publications presenting the main results of tests carried out during the start-up of the first french fast neutron reactor: Rapsodie. The article presents the sodium purification techniques used in the reactor cooling circuits both from the constructional point of view and with respect to results obtained during the first years working. (author) [fr

  20. Sodium purification in Rapsodie; La purification du sodium a Rapsodie

    Energy Technology Data Exchange (ETDEWEB)

    Giraud, B [Commissariat a l' Energie Atomique, Dir. des Piles Atomiques, Cadarache (France). Centre d' Etudes Nucleaires

    1968-07-01

    This report is one of a series of publications presenting the main results of tests carried out during the start-up of the first french fast neutron reactor: Rapsodie. The article presents the sodium purification techniques used in the reactor cooling circuits both from the constructional point of view and with respect to results obtained during the first years working. (author) [French] Ce rapport fait partie d'une serie de publications presentant l'essentiel des resultats des essais effectues a l'occasion du demarrage du premier reacteur francais a neutrons rapides: RAPSODIE. Cet article expose les techniques de la purification du sodium utilise dans les circuits de refroidissement du reacteur tant au point de vue de leur realisation technologique, que des resultats obtenus pendant la premiere annee de fonctionnement. (auteur)