WorldWideScience

Sample records for experimental fast reactor

  1. Introduction of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Matsuba, Ken-ichi; Kawahara, Hirotaka; Aoyama, Takafumi

    2006-01-01

    The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO. (author)

  2. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  3. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

  4. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  5. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    1982-01-01

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

  6. Construction schedule management of China Experimental Fast Reactor

    International Nuclear Information System (INIS)

    Wang Yue

    2012-01-01

    China Experimental Fast Reactor (CEFR) in the first Fast Reactor in China, which is one of large project of the National High Technology Research and Development Program ('863' Program). On 21 st July 2011, CEFR had succeeded to connect to power grid, the target of construction had come true. To a large item, schedule management is one of the most important management, this paper a overall discussion about CEFR item. It has proved that the management of CEFR project is scientific, normative and high-efficiency, it will be valuable for lager Fast Reactor item and designers in interrelated field. (author)

  7. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  8. 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)

  9. Evaluation of fast experimental reactor claddings, (2)

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro; Tanaka, Yasumasa

    1974-01-01

    Thin-walled fine tubes of Type 316 austenitic stainless steel are used for fuel cladding in Joyo (experimental FBR). The material exhibits the change of the mechanical properties in long-time annealing at high temperature, resulting from the precipitation of carbide in structure. In this connection, the experiment and the results on the changes of the microstructure and mechanical properties (proof stress and hardness) are described. The test specimens are the fuel cladding tubes produced for trial for Joyo core and those for FFTF core made in the U.S.A. They were heated between 400 0 and 850 0 C for 1000 hr in vacuum. (Mori, K.)

  10. The SCARABEE experimental fast reactor safety programme already completed

    International Nuclear Information System (INIS)

    Schmitt, A.P.; Teague, H.; Heusener, G.

    1979-08-01

    The SCARABEE in-pile experimental programme comprised a series of tests on unirradiated fuel pins, either single or in seven-pin clusters. The main objective was to obtain information on the mode and consequences of fast reactor fuel pin failure in conditions representative of loss of cooling in a LMFBR. The application of such programmes in full scale reactors leads to the great importance of the interpretation of experimental observations. The interpretation of that programme was carried out jointly by CEA, KFK and UKAEA; this international collaboration led to a sharper focusing on essential features to be modelled in experiments and computer codes and to a valuable convergence of views

  11. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  12. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

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

  13. Project and characteristics of a 5MW experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1986-05-01

    Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 10 15 n/cm 2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author) [pt

  14. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  15. Radiation protection monitoring at the JOYO experimental fast reactor

    International Nuclear Information System (INIS)

    Ouchi, S.; Endo, K.; Susaki, T.

    1979-01-01

    This paper describes the radiation protection monitoring programme for the JOYO experimental fast reactor and some of the health physics problems experienced during the low-power nuclear tests. These include: a detailed description of the centralized radiation monitoring system; the methods and results of the individual monitoring systems; the results of operational monitoring for the handling of new plutonium fuel subassemblies; the evaluation of the external radiation dose rate around the primary coolant system; and the results of an experiment on the thermal dependence of some personnel dose meters. (author)

  16. Theoretical and experimental analysis of fast reactor fuel performance

    International Nuclear Information System (INIS)

    Kummerer, K.R.; Freund, D.; Steiner, H.

    1982-09-01

    In order to predict behavior, performance, and capability of prototypic fuel pins a standard operational scheme for the SNR-300 fast breeder reactor is established considering besides normal operation unscheduled power changes and shutdowns. The behavior during the whole lifetime is calculated using the updated SATURN codes and - for special conditions as power transients and skewed fuel rod power - the new TRANSIENT and TEXDIF codes. The results of these calculations are compared to experimental findings. It is demonstrated that the level of modeling and the knowledge of material properties under irradiation are sufficient for a quantitative description of the fuel pin performance under the above mentioned conditions. (orig.) [de

  17. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  18. Japan: The Experimental Fast Reactor JOYO. Profile 12

    International Nuclear Information System (INIS)

    2017-01-01

    The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

  19. Linear and nonlinear stability analysis, associated to experimental fast reactors

    International Nuclear Information System (INIS)

    Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.

    1980-07-01

    Phenomena associated to the physics of fast neutrons were analysed by linear and nonlinear Kinetics with arbitrary feedback. The theoretical foundations of linear kinetics and transfer functions aiming at the analysis of fast reactors stability, are established. These stability conditions were analitically proposed and investigated by digital and analogic programs. (E.G.) [pt

  20. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.

    1983-01-01

    In order to assure the continued utilization of fission energy, development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed as the best type for future Brazilian nuclear systems. The inherent safety characteristics are superior to current FBRs and an efficient utilization of the abundant thorium is possible. A first step and a basic tool for the development of FBR technologies is the construction and operation of an experimental fast reactor (EFR). A series of core designs for a 90 MW EFR is studied and possible options and the magnitudes of principal parameters are identified. Flexible modifications of the core and sufficiently high fast fluxes for fuel and materials irradiations appear possible. (Author) [pt

  1. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  2. Experimental Methods Related to Coupled Fast-Thermal Systems at the RB Reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2002-01-01

    In addition to the review of RB reactor characteristics this presentation is focused on the coupled fast-thermal systems achieved at the reactor. The following experimental methods are presented: neutron spectra measurements; steady state experiments and kinetic measurements ( β eff ) related to the coupled fast-thermal cores

  3. Primary system thermal-hydraulic simulation of a experimental pool type research fast reactor

    International Nuclear Information System (INIS)

    Borges, E.M.; Braz Filho, F.A.

    1993-01-01

    The first step of the Fast Reactor Program (REARA) is the design of an experimental reactor. To this end a 5 MW t pool type reactor was adapted. The objective of this work is to evaluate the reactor behaviour at the on set protected accidents. The program NALAP was used in this study and the results showed the outstanding safety margins that this reactor type presents inherently. (author)

  4. A study of reactor neutrino monitoring at the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Furuta, H.; Fukuda, Y.; Hara, T.; Haruna, T.; Ishihara, N.; Ishitsuka, M.; Ito, C.; Katsumata, M.; Kawasaki, T.; Konno, T.; Kuze, M.; Maeda, J.; Matsubara, T.; Miyata, H.; Nagasaka, Y.; Nitta, K.; Sakamoto, Y.; Suekane, F.; Sumiyoshi, T.; Tabata, H.

    2012-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3 m from the JOYO reactor core of 140 MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11±1.24(stat.)±0.46(syst.) events/day. Although the statistical significance of the measurement was not enough, backgrounds in such a compact detector at the ground level were studied in detail and MC simulations were found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

  7. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

  8. Aspects of 238Pu production in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Tanaka, Kenya; Itoh, Masahiko; Saito, Masaki

    2005-01-01

    Experimental determination of 238 Pu in 237 Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of 238 Pu production from 237 Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the 238 Pu isotopic ratio. 238 Pu production amount in the irradiated 237 Np samples was determined by a radioanalytical technique. Aspects of 238 Pu production were examined on the basis of the present radioanalysis. The 238 Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu

  9. Linear and nonlinear stability analysis, associated to experimental fast reactors. Part 2

    International Nuclear Information System (INIS)

    Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.

    1980-07-01

    The nonlinear effects in fast reactors kinetics and their stability are studied. The Lyapunov criteria and the Lurie-Letov functions for nonlinear systems were established and simulated. Small oscillations were studied by a Fourier analysis to clarify particular aspects of feedback and load functions in fast reactor at zero power, or/and in normal power level. The results were in agreement with the experimental data existing in the literature. (E.G.) [pt

  10. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  11. Integral test of JENDL dosimetry file using fast neutron field in the Experimental Fast Reactor JOYO

    International Nuclear Information System (INIS)

    Aoyama, Takafumi; Sekine, Takashi

    1999-09-01

    In order to evaluate the applicability of the JENDL dosimetry file, an integral test using a fast neutron spectrum field in the Experimental Fast Reactor JOYO Mark-II core was performed. The dosimeter set consisting of eight reactions of 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 58 Fe(n,γ) 59 Fe, 58 Ni(n,p) 58 Co, 59 Co(n,γ) 60 Co, 63 Cu(n,α) 60 Co, 238 U fission and 237 Np fission was irradiated for approximately 30 days near the core center of the JOYO Mk-II. Neutron flux at the dosimeter position was calculated using the two dimensional discrete ordinate transport code 'DORT'. The core configuration was modeled in XY geometry, and the 100 group cross section set of JSD-J2 / JFT-J2, which was processed from JENDL-2, was utilized. The absolute value of neutron flux was normalized so that the 235 U fission rate using the calculated neutron spectrum agreed with the measured reaction rate. The 103 group cross section data were processed by 'NJOY' code for nuclides to be used in the JOYO dosimetry. As the results of integral test for JENDL/D-99 (new file) and JENDL/D-91 (previous file), calculated values by JENDL/D-99 agreed well with the experimental values, and the C/E ratios ranged from 0.95 to 1.22. By comparing the results between JENDL/D-99 and JENDL/D-91, small differences exist, except for 58 Fe(n, γ) 59 Fe reaction, which was improved significantly in JENDL/D-99. (author)

  12. Pre-design stage of the intermediate heat exchanger for experimental fast reactor

    International Nuclear Information System (INIS)

    Luz, M.; Borges, E.M.; Braz Filho, F.A.; Hirdes, V.R.

    1986-09-01

    This report presents the outlines of a thermal-hydraulic calculation procedure for the pre-design stage of the Intermediate Heat Exchanger for a 5 MW Experimental Fast Reactor (EFR), which can be used in other similar projects, at the same stage of evolution. Heat transfer and heat loss computations for the preliminary design of the heat exchanger are presented. (author) [pt

  13. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  14. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  15. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.; Ninkovic, M.

    1993-01-01

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D 2 O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  16. Construction of fast experimental reactor 'Joyo' from start of construction to criticality

    International Nuclear Information System (INIS)

    Sakata, Hajime

    1977-01-01

    The fast experimental reactor ''Joyo'' is a sodium-cooled, fast neutron reactor using mixed oxide of uranium and plutonium, the first in Japan. The purposes of its construction are to experience and solve the various technical problems expected in the constructions of the prototype reactor ''Monju'' and future practical reactors, and to use as the irradiation facility for developing the fuel and material for fast breeder reactors in Japan after the completion. The construction finished by the end of 1974, and the synthetic functional test was carried out for about two years thereafter. The whole installation was handed over to PNC on March 8, 1977. The reactor attained the criticality on April 24, 1977. The outline of the construction works is described. ''Guidance to the structural design of sodium machinery for Joyo'' was compiled, and the analysis was made according to it. Moreover, various inspection standards regarding welding, electrical machinery, fuel and others were made. The revision of the design for improving the safety and performance was made during the construction at all times. The synthetic functional test was carried out for about two years on 266 items, and subsequently, the criticality test was completed satisfactorily. (Kako, I.)

  17. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    2012-12-01

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  18. Development of the computer code for transient analysis in experimental fast reactor

    International Nuclear Information System (INIS)

    Moreira, M.L.; Sato, E.F.

    1989-01-01

    A calculational model of heat transfer and fluid coolant dynamics, for thermal-hydraulic simulation of the primary system components of a pool type experimental fast breeder reactor, has developed. Programmed in FORTRAN, the SORES code was used to simulate transients as loss of pumping and loss of secondary sodium flow in the EBRII. The SORES results compared with measured data and NATDEMO code results was found to be good. (author) [pt

  19. ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

    International Nuclear Information System (INIS)

    Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

    2008-01-01

    In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

  20. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    International Nuclear Information System (INIS)

    Higinbotham, W.A.

    1994-01-01

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 ± 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF

  1. The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

    Science.gov (United States)

    Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

    2018-01-01

    The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

  2. Fast instrumentation for loss of coolant accident (LOCA) experimental studies pertaining to nuclear reactors

    International Nuclear Information System (INIS)

    Venkat Raj, V.; Sreenivas Rao, G.; Belokar, D.G.; Dolas, P.K.

    1989-01-01

    The loss of coolant accident (LOCA) which involves a breach in the pressure boundary of the primary coolant system (PCS) is one of the postulated accident conditions against which the safety of the reactor system is to be ensured. Mathematical models have been developed to analyse this kind of transients. However, because of the extremely complicated nature of the phenomena involved, it is necessary to validate the analytical models with appropriate experimental data. Many parameters are to be measured during the experiments, out of which temperature, pressure, void fraction and two-phase mass flow rate are the most important parameters. Since the phenomenon is very fast, special fast response instruments are required. This paper deals with the considerations that govern the selection of appropriate instruments and the development of suitable instruments for transient two-phase flow and void fraction measurements. The requirements of the associated fast data acquisition system are also discussed. (author). 4 figs

  3. Numerical analysis of temperature fluctuation in core outlet region of China experimental fast reactor

    International Nuclear Information System (INIS)

    Zhu Huanjun; Xu Yijun

    2014-01-01

    The temperature fluctuation in core outlet region of China Experimental Fast Reactor (CEFR) was numerically simulated by the CFD software Star CCM+. With the core outlet temperatures, flows etc. under rated conditions given as boundary conditions, a 1/4 region model of the reactor core outlet region was established and calculated using LES method for this problem. The analysis results show that while CEFR operates under rated conditions, the temperature fluctuation in lower part of core outlet region is mainly concentrated in area over the edge components (steel components, control rod assembly), and one in upper part is remarkable in area above all the components. The largest fluctuation amplitude is 19 K and the remarkable frequency is below 5 Hz, and it belongs to typically low frequency fluctuation. The conclusion is useful for further experimental work. (authors)

  4. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

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

  5. Development of observation techniques in reactor vessel of experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto

    2010-01-01

    In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1 mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8 mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to

  6. Reprocessing fuel from the Southwest Experimental Fast Oxide Reactor at the Savannah River Plant

    International Nuclear Information System (INIS)

    Gray, L.W.; Campbell, T.G.

    1985-11-01

    The irradiated fuel, reject fuel tubes, and fuel fabrication scrap from the Southwest Experimental Fast Oxide Reactor (SEFOR) were transferred to the Savannah River Plant (SRP) for uranium and plutonium recovery. The unirradiated material was declad and dissolved at SRP; dissolution was accomplished in concentrated nitric acid without the addition of fluoride. The irradiated fuel was declad at Atomics International and repacked in aluminum. The fuel and aluminum cans were dissolved at SRP using nitric acid catalyzed by mercuric nitrate. As this fuel was dissolved in nongeometrically favorable tanks, boron was used as a soluble neutron poison

  7. Hydraulic characteristics of a fast reactor fuel subassembly: An experimental investigation

    International Nuclear Information System (INIS)

    Padmakumar, G.; Velusamy, K.; Prasad, B.V.S.S.; Rajan, K.K.

    2017-01-01

    Highlights: • Fuel subassembly bundle geometry is studied for its hydraulic behaviour. • The results are also compared with data available in literature. • All flow regimes viz. laminar, transition and turbulent is covered for the study. • Pressure drop across different regions of subassembly was also determined. • The effect of external blockage is also studied and reported. - Abstract: Fuel subassemblies of a fast reactor consist of fuel pin bundle with helically wound spacer wires, arranged in a triangular pitch within a hexagonal wrapper. The fuel pins are located within the subassembly. Further the subassembly comprises of a diffuser where the cross section changes from cylindrical to hexagonal, mixing plenum before the exit of pin bundle and a specially designed blockage adapter. Accurate assessment of the pressure drop in the fuel subassembly is essential to ensure adequate core cooling and design of sodium pump. Experimental determination of pressure drop characteristics in the subassembly by simulating the hydraulic condition in the subassemblies of the reactor core is considered essential as a better choice as correlations reported in the literature cannot be directly used for all the complex regions present in the subassembly. This is due to the fact that flows in the interconnecting sections are highly under developed. Further, the flow regime in a fuel subassembly varies from laminar (during shutdown heat removal under natural convection) to completely turbulent under full power condition. To understand the hydraulic characteristics of the 500 MWe Proto type Fast Breeder Reactor (PFBR) fuel subassembly, an experimental facility has been commissioned. Experiments on full scale subassembly with dummy fuel pins have been performed using water as simulant. Experiments have been conducted covering a wide range of Reynolds number encompassing laminar, transition and turbulent regimes. In the rod bundle, no abrupt changes in friction factor were

  8. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  9. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  10. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P. C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  11. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    International Nuclear Information System (INIS)

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  12. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

  13. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  14. Summary report of the experimental fast reactor JOYO MK-III performance test

    International Nuclear Information System (INIS)

    Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro

    2004-03-01

    An upgrading project (MK-III project) was started to improve the irradiation capability of the experimental fast reactor JOYO. In this project, core replacement and increase of the reactor thermal power by the factor 1.4 were necessary for increasing the maximum fast neutron flux by the factor 1.3 and doubling the capacity for irradiation rigs. The modification of the cooling system that included the replacement of the main intermediate heat exchangers and the dump heat exchangers was completed in September 2000. After a series of system function tests, the performance test, of which objective is to fully characterize the upgraded core and heat transfer system, was started in June 2003. Twenty eight tests were selected and carried out as performance test, in order to confirm that the whole plant satisfy the design criteria and have sufficient characteristics (data necessary for safe and steady operation, core management, reactor control and monitoring) as an irradiation bed. After attaining the initial criticality of the core on 2nd July 2003, core characteristics (the excess reactivity, the isotherm temperature reactivity coefficient, the power reactivity coefficient and so on), plant characteristics (the plant heat balance, the adjustment of the temperature control system, the plant behavior at transient), shielding characteristics (dose rate distribution). As the result, it was confirmed that all the criteria regulated was satisfied and the core and plant have sufficient margins for full power operation, which was increased by the factor 1.4. Especially, nuclear analysis accuracy was verified by comparing the calculation with measured core characteristics of the initial core which consists of fifty five fresh fuel subassemblies. The operational data which is supposed to be useful for developing in-core anomaly detection system were also obtained. The operation manual and training simulator and design of next reactor development were revised based on the results

  15. Fast breeder reactor research

    International Nuclear Information System (INIS)

    1975-01-01

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  16. Numerical and on-site experimental dynamic analysis of the Italian PEC fast reactor building

    International Nuclear Information System (INIS)

    Castoldi, A.; Muzzi, F.; Orsi, R.; Panzeri, P.; Pezzoli, P.; Ruggeri, G.; Martelli, A.; Masoni, P.; Brancati, V.

    1988-01-01

    On-site dynamic tests and three-dimensional numerical analysis have been performed by ISMES on behalf of ENEA on the building of the Italian PEC fast reactor test facility. These studies aimed at evaluating the safety margins in the PEC reactor seismic analysis and at providing data for the optimization of the PEC seismic monitoring system. The paper describes the on-site dynamic tests carried out using various excitation methods (two eccentric back-rotating-mass mechanical vibrator, blasting in bore-hole and hydraulic actuators at the building foundations). It highlights the purposes of the four tests campaigns performed at various construction stages and reports the main experimental results. In connection with the experimental tests, a detailed 3D finite element model was set up for fixed base analysis; from the results of the 3D model a simplified equivalent model of the structure was then derived for soil-structure interaction analysis. The mathematical model was validated and calibrated by using the results of the experimental dynamic tests. The main numerical results and the comparisons with the experimental data are presented. (author)

  17. Experimental determination of neutron capture cross sections of fast reactor structure materials integrated in intermediate energy spectra. Vol. 2: description of experimental structure

    International Nuclear Information System (INIS)

    Tassan, S.

    1978-01-01

    A selection of technical documents is given concerning the experimental determination of the neutron capture cross-sections of fast reactor structural materials (Fe, Cr, Ni...) integrated over the intermediate energy spectra. The experimental structure project and modifications of the reactor RB2 for this experiment, together with criticality and safety calculations, are presented

  18. Experimental and numerical analysis of fluid - structure interaction effects in a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Forni, M.; Melloni, R.; Paoluzzi, R.; Bonacina, G.; Castoldi, A.; Zola, M.

    1990-01-01

    Dynamic experiments in air and water (simulating liquid sodium) were performed by ISMES, on behalf of ENEA, on various core element groups of the Italian PEC fast reactor. Bundles of one, seven and nineteen mock-ups reproducing fuel, reflecting and neutron shield elements in full scale were analysed on shaking tables. Tests concerned both groups of equal elements and mixed configurations which corresponded to real core parts. The effects of PEC core-restraint ring were also studied. Seismic excitations of up to 2.5 g were applied to core diagrid. Test results were analysed by use of the one-dimensional program CORALIE and the two-dimensional program CLASH. The study allowed the fluid effects in the PEC core to be evaluated; it also contributed to validation of the above mentioned programs for their general use for fast reactor core analysis. This paper presents the main features of the experimental and the numerical studies and reports comparisons between calculations and measurements. (author)

  19. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

    1965-08-15

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.

  20. The fast breeder reactor

    International Nuclear Information System (INIS)

    Davis, D.A.; Baker, M.A.W.; Hall, R.S.

    1990-01-01

    Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

  1. Fast reactor database

    International Nuclear Information System (INIS)

    1996-02-01

    This publication contains detailed data on liquid metal cooled fast reactors (LMFRs), specifically plant parameters and design details. Each LMFR power plant is characterized by about 400 parameters, by design data and by relevant materials. The report provides general and detailed design characteristics including structural materials, data on experimental, demonstration, prototype and commercial size LMFRs. The focus is on practical issues that are useful to engineers, scientists, managers and university students and professors. The report includes updated information contained in IAEA previous publications on LMFR plant parameters: IWGRF/51 (1985) and IWGFR/80 (1991) and reflects experience gained from two consultants meetings held in Vienna (1993,1994). This compilation of data was produced by members of the IAEA International Working Group on Fast Reactors (IWGFR)

  2. Basic principles of accounting and control of nuclear materials in the BOR-60 experimental fast reactor

    International Nuclear Information System (INIS)

    Gryazev, V.M.; Gadzhiev, G.I.; Alekseev, I.N.

    1979-01-01

    Under a contract with the International Atomic Energy Agency, the V.I. Lenin Atomic Reactor Research Institute is currently carrying out a study of ways of organizing a nuclear materials accounting and control system for the BOR-60 fast reactor. Some results of this study are presented in the paper. The special physical and technological features of fast reactors create additional difficulties in safeguards systems and give rise to a number of new possibilities for the illicit removal of nuclear materials. These questions are discussed with reference to the BOR-60 reactor but the conclusions are probably applicable to all fast reactors. The proposed accounting and control system is based on non-destructive measurements of the amount of fissile materials in the operating fuel assemblies and screened bundles of the reactor, on the independent control of the principal facility parameters (a list of which is given) and on an automated information collection and evaluation system. Visual means of inspection can be very effective in fast reactor safeguards systems, especially for controlling storage, but they are not used with the BOR-60 reactor. (author)

  3. Determination of hydrazine in third loops of China experimental fast reactor by spectrophotometry

    International Nuclear Information System (INIS)

    Huang Wenjie; Wang Mi; Gao Yaopeng; Xie Chun; Yu Xiaochen

    2013-01-01

    The method for the determination of hydrazine by Uv-vis spectrophotometer was proposed. The coloration conditions and instrument parameters were also optimized. In HCl, hydrazine formed a yellow azine with para-dimethyl aminobenzaldehyde ((CH 3 ) 2 NC 6 H 4 CHO), and then determined by spectrophotometer. The complex's maximum absorption was exhibited at 458 nm. The coloration effect was excellent in conditions of 1% HCl, 10 mL para-dimethyl aminobenzaldehyde and 10 minutes' developing time. A good liner relationship was obtained in the range of 5∼200 μg/L, and the recovery was (101.1±1.9)%. This method was used in the third loop of China experimental fast reactor with satisfactory results. (authors)

  4. Specialist committee's review reports for experimental fast reactor JOYO' MK-III performance tests

    International Nuclear Information System (INIS)

    Yamashita, Kiyonobu; Okubo, Toshiyuki; Kamide, Hideki

    2004-02-01

    Performance tests (startup-physics tests and power elevation tests) were planed for experimental fast reactor 'JOYO' MK-III where irradiation performances were upgraded by power increase from 100 to 140 MW. The reactor safety committee of O-arai Engineering Center has established a specialist committee for 'JOYO' MK-III Performance Tests at the first meeting of 2003 on 23th. April 2003, to accomplish the tests successfully. Subjects of the specialist committee were reviews of following items covering a wide range. 1) Contents of modification works. 2) Reflections of functional test results to the plant and facilities. 3) Reflections of safety rule modification to instruction and manual for operation. 4) Quality assurances and pre-calculation for performance test. 5) Inspection plan and its results. 6) Adequacy of performance test plan. 7) Confirmation of performance test results. Before test-starts, the specialist committee has confirmed by reviewing the items from 1) to 6) based on explanations and documents of the Division of Experimental Reactor, that the test plan and pre-inspections are adequate. After the tests, the specialist committee had confirmed by reviewing the item 7) in the same way, that the each test result satisfies the corresponding criterion. The specialist committee has concluded from these review's results before and after the tests that the 'JOYO' MK-III Performance Tests were carried out appropriately. Besides, the first criticality of the JOYO MK-III was achieved on 2nd. July 2003, and the continuous full power operation was carried on 20th. Nov. 2003. Finally, all performance tests were completed by the pass of the last governmental pre-serviced inspection (dose rate measurement during the shut down condition). (author)

  5. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  6. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  7. Utilization of the experimental reactor Osiris for the study and the development of fuels of the fast neutron reactor type

    International Nuclear Information System (INIS)

    Marcon, M.; Faugere, J.L.; Genthon, J.P.; Maillot, R.

    1977-01-01

    Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O 2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation [fr

  8. Some study on radiation resistance and reliability of piston ring of waste gas compressor for fast breeder experimental reactor

    International Nuclear Information System (INIS)

    Muramatsu, Takio; Hidaka, Tsukasa

    1976-01-01

    In the fast breeder experimental reactor ''Joyo'', the gaseous wastes such as reactor cover argon, reactor seal nitrogen gas, fuel handling waste gas etc. shall be collected, compressed and storaged for decaying their activity. Compressors applied in the above process have new type oilless piston rings of Teflon filled with graphite, which might be affected by radioactivity of the waste gases. This report deals with some study on the gamma iradiation effects on the plastic piston rings such as tensile strength, elongation, shock and hardness effects under several irradiation doses and on durability test of the irradiated piston rings under the same compression ratio. (auth.)

  9. New experimental space for irradiating samples by RA reactor fast neutron flux at temperatures up to 100 deg C

    International Nuclear Information System (INIS)

    Pavicevic, M.; Novakovic, M.; Zecevic, V.

    1961-01-01

    The objective of this paper is to present adaptation of the RA reactor which would enable samples irradiation by fast neutrons and describe new experimental possibilities. New experimental space was achieved using hollow fuel elements which have been reconstructed to enable placement of irradiation capsules inside the tube. This paper includes thermal analysis and describes problems related to operation, safety and radiation protection issues which arise from using reconstructed fuel elements

  10. Simulation research about China Experimental Fast Reactor steam turboset based on Flowmaster platform

    International Nuclear Information System (INIS)

    Yan Hao; Tian Zhaofei

    2014-01-01

    In the third loop of China Experimental Fast Reactor (CEFR), steam turboset take an important role in converting heat energy into electric energy. However, turbo sets have not been operated on the condition of more than 40%P_0 (P_0 is full power) since they were installed. Thus it is necessary to make an analogue simulation. Based on the real models of turbo sets in CEFR, simulation models were created with the help of Flowmaster platform. By using such simulation models, a steady state result in full power circumstance was got, which is in accordance with design parameters. Meanwhile, a transient state simulation with operating condition ranging from full power to 40%P_0 was accomplished and a result which verifies part of performance and running conditions of turbo sets was got. The result of analogue simulation shows that based on Flowmaster platform, the running condition of simulation models can comply with design requirement, and offer reference values to the actual running. Such simulation models can also offer reference values to other simulation models in the third loop of CEFR. (authors)

  11. Fast neutron flux in the RA reactor experimental channels; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Dobrosavljevic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Fast neutron flux in the RA reactor experimental channels was determined by using threshold reaction detectors. The (n,p) type reactions S{sub 32} (n,p)P{sub 32}, and Al{sub 24} (n,p)Na{sub 24}. Prepared sulphur and phosphorous foils were placed in cadmium boxes and irradiated in experimental channels VK-5, VK-7 and VK-9. Gold foils were irradiated simultaneously for controlling the reactor power. Reactor power was 100 kW during irradiation of half an hour. Activity of P{sub 32} and S{sub 31} after reactor shutdown was measured by 4{pi} counter and three calibrated GM counters. Absolute neutron flux was determined by using thus obtained data.

  12. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  13. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  14. Fast reactor database. 2006 update

    International Nuclear Information System (INIS)

    2006-12-01

    Liquid metal cooled fast reactors (LMFRs) have been under development for about 50 years. Ten experimental fast reactors and six prototype and commercial size fast reactor plants have been constructed and operated. In many cases, the overall experience with LMFRs has been rather good, with the reactors themselves and also the various components showing remarkable performances, well in accordance with the design expectations. The fast reactor system has also been shown to have very attractive safety characteristics, resulting to a large extent from the fact that the fast reactor is a low pressure system with large thermal inertia and negative power and temperature coefficients. In addition to the LMFRs that have been constructed and operated, more than ten advanced LMFR projects have been developed, and the latest designs are now close to achieving economic competitivity with other reactor types. In the current world economic climate, the introduction of a new nuclear energy system based on the LMFR may not be considered by utilities as a near future option when compared to other potential power plants. However, there is a strong agreement between experts in the nuclear energy field that, for sustainability reasons, long term development of nuclear power as a part of the world's future energy mix will require the fast reactor technology, and that, given the decline in fast reactor development projects, data retrieval and knowledge preservation efforts in this area are of particular importance. This publication contains detailed design data and main operational data on experimental, prototype, demonstration, and commercial size LMFRs. Each LMFR plant is characterized by about 500 parameters: physics, thermohydraulics, thermomechanics, by design and technical data, and by relevant sketches. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors with complete technical information of a total of 37 LMFR

  15. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Lineberry, M.J.

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs

  16. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics

  17. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  18. Theoretical and experimental investigation of the nonlinear structural dynamics of Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Liebe, R.

    1978-04-01

    This study describes theoretical and experimental investigations of the dynamic deformation behavior of single and clustered fuel elements under local fault conditions in a Fast Breeder Reactor core. In particular an energetic molten-fuel-coolant-interaction (FCI) is assumed in one subassembly with corresponding pressure pulses, which may rupture the wrapper and load the adjacent fuel elements impulsively. Associated coherent structural deformation may exceed tolerable and damage the control rods. To attack the outlined coupled fluid-structure-interaction problem it is assumed, that the loading at the structures is known in space and time, and that there is no feedback from the deformation response. Then current FCI-knowledge and experience from underwater core model explosion tests is utilized to estimate upper limits of relevant pulse characteristics. As a first step the static carrying capacity of the rigid-plastic hexagonal wrapper tube is calculated using the methods of limit analysis. Then for a general dynamic simulation of the complete elastoplastic subassembly response the concept of a discrete nonlinear hinge is introduced. A corresponding physical lumped parameter hinge model is presented, and general equations of motion are derived using D'Alembert's principle. Application to the static and dynamic analysis of a single complete fuel element includes the semiempirical modelling of the fuel-pin bundle by a homogeneous compressible medium. Most important conclusions are concerning the capability of the theoretical models, the failure modes and threshold load levels of single as well as clustered SNR-300 fuel elements and the safety relevant finding, that only limited deformations are found in the first row around the incident element. This shows in agreement with explosion test results that the structured and closely spaced fuel elements constitute an effective, inherent barrier against extreme dynamic loadings. (orig.) [de

  19. Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    2007-01-01

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

  20. Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

    International Nuclear Information System (INIS)

    Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi

    2002-01-01

    In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

  1. Design and adjustment on test bed of replacing subassembly machine control system for China experimental fast reactor

    International Nuclear Information System (INIS)

    Dong Shengguo; Ma Hongsheng; Zhao Lixia

    2008-01-01

    The present research concerns in the design and adjustment of replacing sub- assembly machine control system of China Experimental Fast Reactor. The design of replacing subassembly machine control system adopts some electric equipments, such as programmable controllers, digital DC drivers. The designed control system was adjusted on the test bed. The results indicate that the operation of the control system is steady and reliable, and designed control system can meet the needs of the design specification. (authors)

  2. Fast breeder reactors

    International Nuclear Information System (INIS)

    Waltar, A.E.; Reynolds, A.B.

    1981-01-01

    This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of reactor theory, heat transfer, and fluid mechanics. The book is expected to be used most widely for a one-semester general course on fast breeder reactors, with the extent of material covered to vary according to the interest of the instructor. The book could also be used effectively for a two-quarter or a two-semester course. In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety. Methodology in fast reactor design and analysis, together with physical descriptions of systems, is emphasized in this text more than numerical results. Analytical and design results continue to change with the ongoing evolution of FBR design whereas many design methods have remained fundamentally unchanged for a considerable time

  3. The fast breeder reactor

    International Nuclear Information System (INIS)

    Patterson, W.

    1990-01-01

    The author criticises the United Kingdom Atomic Energy Authority's fast breeder reactor programme in his evidence to the House of Commons Select Committee on Energy in January 1990. He argues for power generation by renewable means and greater efficiency in the use rather than in the generation of electricity. He refutes the arguments for nuclear power on the basis of reduced global warming as he claims support technology produces significant amounts of carbon dioxide in any case. Serious doubts are raised about the costs of a fast breeder reactor programme compared to, say, generation by pressurised water reactors. The idea of a uranium scarcity in several decades is also refuted. The reliability of fast breeder reactor technology is called into question. He argues against reprocessing plutonium for economic, health and safety reasons. (UK)

  4. Comparison of diffusion and transport theory analysis with experimental results in fast breeder test reactor

    International Nuclear Information System (INIS)

    Sathyabama, N.; Mohanakrishnan, P.; Lee, S.M.

    1994-01-01

    A systematic analysis has been performed by 3 dimensional diffusion and transport methods to calculate the measured control rod worths and subassembly wise power distribution in fast breeder test reactor. Geometry corrections (rectangular to hexagonal and diffusion to transport corrections are estimated for multiplication factors and control rod worths. Calculated control rod worths by diffusion and transport theory are nearly the same and 10% above measured values. Power distribution in the core periphery is over predicted (15%) by diffusion theory. But, this over prediction reduces to 8% by use of the S N method. (authors). 9 refs., 4 tabs., 3 fig

  5. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

    International Nuclear Information System (INIS)

    Ushiki, Hiroshi; Ito, Hiromichi; Okuda, Eiji; Suzuki, Nobuhiro; Sasaki, Jun; Oota, Katsu; Kawahara, Hirotaka; Takamatsu, Misao; Nagai, Akinori; Okawa, Toshikatsu

    2015-01-01

    In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

  6. The fast breeder reactor

    International Nuclear Information System (INIS)

    Keck, O.

    1984-01-01

    Nowadays the fast-breeder reactor is a negative symbol of advanced technology which is getting out of control and, due to its complexity, is incomprehensible for politicians and therefore by-passes the established order. The author lists the most important decisions over state aid to the fast-breeder-reactors up until the mid-seventies and uses documents from the appropriate advisory bodies as reference. He was also aided by interviews with those directly involved with the project. The empirical facts forces us to discard our traditional view of the relationship between state and industry with regard to advanced technology. The author explains that it is impossible to find any economic value in the fast-breeder reactor. The insight gained through this project allows him to draw conclusions which apply to all aspects of state aid to advanced technology. (orig.) [de

  7. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1982-01-01

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  8. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E; Andersson, T L; Brunfelter, B; Kockum, J; Londen, S O; Tiren, L I

    1965-12-15

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

  9. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

    2014-01-01

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  10. 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.

  11. An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Maeda, Shigetaka; Mitsugashira, Toshiaki

    2005-01-01

    Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the 241 Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the 240 Pu to 239 Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor

  12. Integral Fast Reactor concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path

  13. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  14. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  15. The instrumentation of fast reactor

    International Nuclear Information System (INIS)

    Endo, Akira

    2003-03-01

    The author has been engaged in the development of fast reactors over the last 30 years with both an involvement with the early technology development on the experimental breeder reactor Joyo, and latterly continuing this work on the prototype breeder reactor, Monju. In order to pass on this experience to younger engineers this paper is produced to outline this experience in the sincere hope that the information given will be utilised in future educational training material. The paper discusses the wide diversity on the associated instrument technology which the fast breeder reactor requires. The first chapter outlines the fast reactor system, followed by discussions on reactor instrumentation, measurement principles, temperature dependencies, and verification response characteristics from various viewpoints, are discussed in chapters two and three. The important issues of failed fuel location detection, and sodium leak detection from steam generators are discussed in chapters 4 and 5 respectively. Appended to this report is an explanation on the methods of measuring response characteristics on instrumentation systems using error analysis, random signal theory and measuring method of response characteristic by AR (autoregressive) model on which it appears is becoming an indispensable problem for persons involved with this technology in the future. (author)

  16. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  17. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1985-01-01

    During the past two years, scientists from Argonne have developed an advanced breeder reactor with a closed self contained fuel cycle. The Integral Fast Reactor (IFR) is a new reactor concept, adaptable to a variety of designs, that is based on a fuel cycle radically different from the CRBR line of breeder development. The essential features of the IFR are metal fuel, pool layout, and pyro- and electro-reprocessing in a facility integral with the reactor plant. The IFR shows promise to provide an inexhaustible, safe, economic, environmentally acceptable, and diversion resistant source of nuclear power. It shows potential for major improvement in all of the areas that have led to concern about nuclear power

  18. Experimental study of the attenuation waves oriented to transients caused by the sodium-water explosive reaction in fast reactors

    International Nuclear Information System (INIS)

    Pedroso, L.J.

    1990-01-01

    One of the problems related to fluid-structure interaction that can compromise the structural integrity of components of a fast reactor is the explosion caused by the sodium-water reaction, in the case of a flood at the level of the thermic exchange wall at the steam generator. In this paper we have considered the aspects of the pressure-waves damping caused by the reaction, when these waves transverse certain perforated structures. In order to solve this problem, we also adopted a parametric experimental approach, using a scale model (RIO test rig). (author)

  19. Dounreay fast reactor

    International Nuclear Information System (INIS)

    Maclennan, R.; Eggar, T.; Skeet, T.

    1992-01-01

    The short debate which followed a private notice question asking for a statement on Government policy on the future of the European fast breeder nuclear research programme is reported verbatim. In response to the question, the Minister for Energy said that the Government had decided in 1988 that the Dounreay prototype fast reactor would close in 1994. That decision had been confirmed. Funding of fast breeder research and development beyond 1993 is not a priority as commercialization is not expected until well into the next century. Dounreay will be supported financially until 1994 and then for its subsequent decommissioning and reprocessing of spent fuel. The debate raised issues such as Britain losing its lead in fast breeder research, loss of jobs and the Government's nuclear policy in general. However, the Government's position was that the research had reached a stage where it could be left and returned to in the future. (UK)

  20. Advanced Safeguards Approaches for New Fast Reactors

    International Nuclear Information System (INIS)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-01-01

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to 'breed' nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and 'burn' actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is 'fertile' or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing 'TRU'-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II 'EBR-II' at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line--a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors

  1. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  2. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    1983-02-01

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  3. Gas-cooled fast reactors. Motivation and presentation of the ENIGMA program in the MASURCA experimental critical facility

    International Nuclear Information System (INIS)

    Tommasi, Jean; Jacqmin, Robert; Mellier, Frederic

    2005-01-01

    This paper describes a new experimental physics program in support of gas cooled fast reactor (GCFR) design studies, called ENIGMA, to be performed in the MASURCA critical facility at CEA-Cadarache, France. The prospective GCFR design studies at CEA are presented, as well as the specific neutronics features needing an extension of the validation of calculation tools and nuclear data. The relevant existing experiments are briefly reviewed and the need for new experimental data is pointed out. The first phase of the proposed new experiments includes a reference core with a representative spectrum, and a series of central core substitutions involving spectrum shifts, streaming studies, low-grade Pu substitutions, innovative material (Si, Zr) substitutions. Reflector substitution zones will include elements foreseen for the reflectors (Si, Zr, C). Subsequent phases will involve larger amounts of low-grade Pu or innovative materials, and configurations representative of experimental and demonstration GCFRs. (author)

  4. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

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

  5. Knowledge management in fast reactors

    International Nuclear Information System (INIS)

    Kuriakose, K.K.; Satya Murty, S.A.V.; Swaminathan, P.; Raj, Baldev

    2010-01-01

    This paper highlights the work that is being carried out in Knowledge Management of Fast Reactors at Indira Gandhi Centre for Atomic Research (IGCAR) including a few examples of how the knowledge acquired because of various incidents in the initial years has been utilized for the successful operation of Fast Breeder Test Reactor. It also briefly refers to the features of the IAEA initiative on the preservation of Knowledge in the area of Fast Reactors in the form of 'Fast Reactor Knowledge Organization System' (FR-KOS), which is based on a taxonomy for storage and mining of Fast Reactor Knowledge. (author)

  6. A review of theoretical and experimental studies underlying the thermal-hydraulic design of fast reactor fuel elements

    International Nuclear Information System (INIS)

    McAreavey, C.G.; Betts, C.

    1979-01-01

    The economic performance of fast reactors is closely linked to the achievable burn-up of heavy atoms, that is to the endurance life of the fuel pins. The safety case must also be concerned with the integrity of the cladding, since this is the primary containment envelope for fission products. It is thus important to ensure that cladding temperatures during reactor operation are limited to levels which incur no serious impairment of mechanical properties. The function of thermal-hydraulic analysis is to provide fuel element designers with the means of achieving this objective. This paper reviews the theoretical approaches which have been developed and applied in the UK in the design of LMFBR fuel and breeder sub-assemblies, control rods and experimental clusters. It also presents results of experimental studies undertaken to develop a better understanding of coolant flow distribution and mixing problems in these components, and to provide essential data for computer codes. Problem areas in this field are highlighted, particularly the difficulties arising due to irradiation induced distortions. Reference is made to the experimental and theoretical developments which are in progress, or may be required, to provide adequate predictions of fuel pin temperatures at high burn-up. (author)

  7. Fast breeder reactors

    International Nuclear Information System (INIS)

    Ollier, J.L.

    1987-01-01

    The first industrial-scale fast breeder reactor (FBR) is the Superphenix I at Crays-Melville. It was designed and built by Novatome, a French company, and Ansaldo, an Italian company. The advantages of FBRs are summarized. The status of Superphenix and the testing schedule is given. The stages in its power escalation in 1986 are given. The article is optimistic about the future for FBRs and expects FBRs to take over from PWRs at the beginning of the 21st Century. To achieve economic viability, European financial cooperation for the research and development programme is advocated. (UK)

  8. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

    2015-01-01

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

  9. New fast reactor installation concept

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The large size and complexity of fast reactor installations are emphasised and these difficulties will be increased with the advent of fast reactors of higher power. In this connection a new concept of fast reactor installation is described with a view to reducing the size of the installation and enabling most components, including even the primary vessel, to be constructed within the confines of a workshop. Full constructional details are given. (U.K.)

  10. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

    Orlov, V.V.

    1997-01-01

    1.Three aspects of nonproliferation relevant to nuclear power are: Pu buildup in NPP spent fuel cooling ponds (∼ 104 t in case of consumption of ∼ 107 t cheap uranium). Danger of illegal radiochemical extraction of Pu for weapons production; Pu extraction from NPP fuel at the plants available in nuclear countries, its burning along with weapon-grade Pu in NPP reactors or in special-purpose burners; increased hazard of nuclear weapons sprawl with breeders and closed fuel cycle technology spreading all over the world. 2.The latter is one of major obstacles to creation of large-scale nuclear power. 3.Nuclear power of the first stage using 235 U will be able to meet the demands of certain fuel-deficient countries and regions, replacing ∼ 5-10% of conventional fuels in the global consumption for a number of decades. 4.Fast reactors of the first generation and the currently employed fuel technology are far from exhausting their potential for solving economic problems and meeting the challenges of safety, radioactive waste and nonproliferation. Development of large-scale nuclear power will become an option accepted by society for solving energy problems in the following century, provided a breeder technology is elaborated and demonstrated in the next 15-20 years, which would comply with the totality of the following requirement: full internal Pu breeding deterministic elimination of severe accidents involving fuel damage and high radioactivity releases: fast runaway, loss of coolant, fires, steam and hydrogen explosions, etc.; reaching a balance between radioactive wastes disposed of and uranium mined in terms of radiation hazard; technology of closed fuel cycle preventing its use for Pu extraction and permitting physical protection from fuel thefts;economic competitiveness of nuclear power for most of countries and regions, i.e. primarily the cost of NPPs with fat reactors is to be below the cost of modern LWR plants, etc

  11. Experimental investigation into fast pyrolysis of biomass using an entrained flow reactor

    Science.gov (United States)

    Bohn, M.; Benham, C.

    1981-02-01

    Pyrolysis experiments were performed with steam as a carrier gas and two different feedstocks - wheat straw and powdered material derived from municipal solid waste (ECO-II TM). Reactor wall temperature was varied from 7000 to 1400 C. Gas composition data from the ECO-II tests were comparable to previously reported data but ethylene yield appeared to vary with reactor wall temperature and residence time. The important conclusion from the wheat straw tests is that olefin yields are about one half that obtained from ECO-II. Evidence was found that high olefin yields from ECO-II are due to the presence of plastics in the feedstock.

  12. Scientific and technical conference Thermophysical experimental and calculating and theoretical studies to justify characteristics and safety of fast reactors. Thermophysics-2012. Book of abstracts

    International Nuclear Information System (INIS)

    Kalyakin, S.G.; Kukharchuk, O.F.; Sorokin, A.P.

    2012-01-01

    The collection includes abstracts of reports of scientific and technical conference Thermophysics-2012 which has taken place on October 24-26, 2012 in Obninsk. In abstracts the following questions are considered: experimental and calculating and theoretical studies of thermal hydraulics of liquid-metal cooled fast reactors to justify their characteristics and safety; physico-chemical processes in the systems with liquid-metal coolants (LMC); physico-chemical characteristics and thermophysical properties of LMC; development of models, computational methods and calculational codes for simulating processes of of hydrodynamics, heat and mass transfer, including impurities mass transfer in the systems with LMC; methods and means for control of composition and condition of LMC in fast reactor circuits on impurities and purification from them; apparatuses, equipment and technological processes at the work with LMC taking into account the ecology, including fast reactors decommissioning; measuring techniques, sensors and devices for experimental studies of heat and mass transfer in the systems with LMC [ru

  13. Fast reactor programme

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1976-11-01

    Estimated reactivity effects of fission products in the SNR-300 fast breeder are given. Neutron cross sections of 127 I and 129 I are also given. Results of the in-pile canning failure experiments on fuel pins R54-F35 and F39 are discussed. Sinter experiments using mixed UC-UN powders are reported. Results of tensile tests on high-dose and low-dose irradiated specimens of 18Cr1 1Ni stainless steel (DIN 1.4948) used in the SNR-300 reactor vessel are given. It is shown that the aerosol behaviour in condensing sodium vapour can be described by the same MADCA model developed for the decay of aerosols in condensing water vapour. Results of heat transfer measurements in the electrically heated 28-rod bundle under liquid-phase and subsequently under two-phase conditions are commented on

  14. Uranium and the fast reactor

    International Nuclear Information System (INIS)

    Price, T.

    1982-01-01

    The influence of uranium availability upon the future of the fast reactor is reviewed. The important issues considered are uranium reserves and resources, uranium market prices, fast reactor economics and the political availability of uranium to customers in other countries. (U.K.)

  15. The safety of fast reactors

    International Nuclear Information System (INIS)

    Justin, F.

    1976-01-01

    A response is made to the main questions that a man in the street may arise concerning fast breeder reactors, in particular: the advantages of this line, dangerous materials contained in fast breeder reactors, containment shells protecting the environment from radiations, main studies now in progress [fr

  16. Periodic safety review of the experimental fast reactor JOYO. Review of the activity for safety

    International Nuclear Information System (INIS)

    Maeda, Yukimoto; Kashimura, Youichi; Suzuki, Toshiaki; Isozaki, Kazunori; Hoshiba, Hideaki; Kitamura, Ryoichi; Nakano, Tomoyuki; Takamatsu, Misao; Sekine, Takashi

    2005-02-01

    Periodic safety review (Review of the activity for safety) which consisted of 'Comprehensive evaluation of operation experience' and Incorporation of the latest technical knowledge' was carried out up to January 2005. 1. Comprehensive evaluation of operation experience. It was confirmed that the effectual activities for safety through the operation of JOYO were carried out in terms of (1) Operation management, (2) Maintenance management, (3) Fuel management, (4) Radiation management, (5) Radioactive waste management, (6) Emergency planning and (7) Feedback of incidents and failures. 2. Reflection of the latest technical knowledge. It was confirmed that the latest technical knowledge including regulation and guide line established by Nuclear Safety Commission of Japan until March 31st. 2003 were properly reflected in impressing the safety of the reactor. As a result, it was evaluated that the activity for safety was carried out effectually, and no additional measure was identified continual safe operation of the reactor. (author)

  17. Fast reactor physics - an overview

    International Nuclear Information System (INIS)

    Lee, S.M.

    2004-01-01

    An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

  18. Homogenization of the coefficient of diffusion: influence of modelling and of the laplacian for fast power reactors and experimental mockups

    International Nuclear Information System (INIS)

    Gho, C.J.

    1984-10-01

    Neutron transport calculation of reactors is based on the definition of homogeneized cell constants, the diffusion coefficient among others. The formalism of the evaluation of the diffusion coefficient, as also the cell model used may introduced uncertainties in results. The present study allowed to estimate these uncertainties in the case of fast neutron power reactors and criticical mockups. The validation of new simple methods and the definition of references is a consequence of this work [fr

  19. Fast mixed spectrum reactor concept

    International Nuclear Information System (INIS)

    Kouts, H.J.C.; Fischer, G.J.; Cerbone, R.J.

    1979-04-01

    The Fast Mixed Spectrum Reactor is a highly promising concept for a fast reactor with improved features of proliferation resistance, and excellent utilization of uranium resources. In technology, it can be considered to be a branch of fast breeder development, though its operation and implications are different from those of FBR'S in important respects. Successful development programs are required in several areas to bring FMSR to reality, but the payoff from a successful program can be high

  20. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1981-06-01

    The accuracy requirements and the status of the evaluated fission-product cross sections for fast reactors are reviewed; the work on calculating the sensitivity of the sodium void effect to fission-product cross sections is described; some results of the intercomparison of adjusted data sets for capture cross sections of fission-products (RCN-2A and CARNAVAL-IV) are discussed; the applicability of the maximum-likelihood method for the analysis of resolved resonance parameters for a large class of fission-product nuclides is demonstrated; the neutron cross sections for corrosion product 64 Ni are evaluated. Some results of post-irradiation examination of a loss-of-cooling experiment are given; the progress in testing the equipment and instrumentation for transient-overpower experiments is reported. The proceedings in the thermochemical investigations on uranium compounds with some fission-products are described. The creep behaviour of a heat of DIN 1.4948 parent metal is investigated with respect to the changes in strain with different test temperatures. Sodium smoke aerosols have been produced and analysed with respect to their aerodynamic behaviour and morphology. The two-phase local boiling experiments have been analysed to find criteria for the occurrence of different boiling regimes with the objection to deduce general dryout correlations

  1. Fast reactor recharging device

    International Nuclear Information System (INIS)

    Artemiev, L.N.; Kurilkin, V.V.

    1979-01-01

    Disclosure is made of a device for recharging a fast-neutron reactor, intended for the transfer of fuel assemblies and rods of the control and safety system, having profiled heads to be gripped on the outside. The device comprises storage drums whose compartments for rods of the control and safety system are identical to compartments for fuel assemblies. In order to store and transport rods of the control and safety system from the storage drums to the recharging mechanism provision is made for sleeve-type holders. When placed in such a holder, the dimensions of a rod of the control and safety system are equal to those of a fuel assembly. To join a holder to a rod of the control and safety system, on the open end of each holder there is mounted a collet, whereas on the surface of each rod of the control and safety system, close to its head, there is provided an encircling groove to interact with the collet. The grip of the recharging mechanism is provided with a stop interacting with the collet in order to open the latter and withdraw the safety and control system rod from its holder

  2. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  3. Fast reactor research activities in Brazil

    International Nuclear Information System (INIS)

    Menezes, A.

    1998-01-01

    Fast reactor activities in Brazil have the objective of establishing a consistent knowledge basis which can serve as a support for a future transitions to the activities more directly related to design, construction and operation of an experimental fast reactor, although its materialization is still far from being decided. Due to the present economic difficulties and uncertainties, the program is modest and all efforts have been directed towards its consolidation, based on the understanding that this class of reactors will play an important role in the future and Brazil needs to be minimally prepared. The text describes the present status of those activities, emphasizing the main progress made in 1996. (author)

  4. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

    2010-01-01

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  5. Fast reactor collaboration in Europe

    International Nuclear Information System (INIS)

    Smith, G.E.I.

    1987-01-01

    Fast reactors have been developed in several European countries, the United Kingdom, France, Germany and Italy. A suggestion to collaborate on fast reactor research and development resulted in an Intergovernmental Memorandum of Understanding signed in 1984 by the UK, France, Germany, Italy and Belgium. Holland was expected to join later. This provided for co-operation between electric utilities, reactor design, research and development companies and fuel cycle companies. Three steering committees have so far been set up, the European fast reactor utilities Group, the European research and development and the European fuel cycle steering committees. Progress on these is detailed. The main areas of technology exchange are listed in the Appendix. The possibility exists for a series of three large demonstration plants to be built in Europe and a fuel reprocessing plant to confirm the reactor system. (U.K.)

  6. The integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Marchaterre, J.F.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

  7. Review of fast reactor activities

    International Nuclear Information System (INIS)

    1982-01-01

    A description of some highlights of the activities performed by the Commission of the European Communities in the field of fast reactors is given. They fall into two categories: coordinating and harmonizing activities and research activities. The former are essentially performed in the frame of the Fast Reactor Coordinating Committee (FRCC), the latter in the Commission's Joint Research Center and to some extent under contract in research centers of the Member States

  8. Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2016-05-15

    Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

  9. Interfacial effects in fast reactors

    International Nuclear Information System (INIS)

    Saidi, M.S.; Driscoll, M.J.

    1979-05-01

    The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed to measure U-238 capture rates near th blanket--reflector interface in the MIT Blanket Test Facility. Prior MIT experiments on a thorium--uranium interface in a blanket assembly were also reanalyzed. Extremely localized fertile capture rate increases of on the order of 50% were measured immediately at the interfaces relative to extrapolation of asymptotic interior traverses, and relative to state-of-the-art (LIB-IV, SPHINX, ANISN/2DB) calculations which employ infinite-medium self-shielding throughout a given zone. A method was developed to compute a spatially varying background scattering cross section per absorber nucleus which takes into account both homogeneous and heterogeneous effects on the interface flux transient

  10. Review of fast reactor activities

    International Nuclear Information System (INIS)

    Haeussermann, W.; Royen, J.

    1978-01-01

    Since 1971, when the Co-ordinating Group on Gas-Cooled Fast reactors Development was set up, the participating countries have maintained an interest in keeping this option as a back-up solution to the sodium cooled fast reactors. Two different concepts were investigated, one based on coated particle type fuel elements and the other on pin type fuel elements. The coated particles studies have been brought to an end, and resources were concentrated on the further development of the pin type concept. The work done in previous years covered design and safety investigations, heat transfer studies and irradiation experiments in thermal reactors

  11. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1982-01-01

    A review of fast reactor activities in India is introduced. One stage of construction of the Fast Breeder Test Reactor (FBTR) and design studies for 500MWe Prototype Fast Breeder Reactor (PFBR) are briefly summarized. The emphasis is on fast reactor physics, materials studies, radiochemistry, and the safety and fuel reprocessing programme

  12. Fast breeder reactor

    International Nuclear Information System (INIS)

    Ito, Shin-ichi; Maki, Koichi.

    1975-01-01

    Object: To conserve loaded fuel, aquire controllable surplus reaction degree, increase the breeding index, flatten output and improve sealing of neutrons by inserting a decelerating substance in a blanket section. Structure: A decelerating substance such as beryllium or beryllium oxide is inserted in a blanket section between an outer reactor core and reflector. With this arrangement, neutrons are decelerated to increase the low energy components, which are partly subjected to reflection by the outer reactor core to thereby reduce leakage of neutrons from the reactor core. (Kamimura, M.)

  13. Astrid (fast breeder nuclear reactor)

    International Nuclear Information System (INIS)

    2014-01-01

    This document presents ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a French project of sodium-cooled fast breeder reactor, fourth generation reactor which should be fuelled by uranium 238 rather than uranium 235, and should therefore need less extracted natural uranium to produce electricity. The operation principle of fast breeder reactors is described. They notably directly consume plutonium, allow an easier radioactive waste management as they transform long life radioactive elements into shorter life elements by transmutation. The regeneration process is briefly described, and the various operation modes are evoked (iso-generator, sub-generator, and breeder). Some peculiarities of sodium-cooled reactors are outlined. The Astrid operation principle is described, its main design innovations outlined. Various challenges are discussed regarding safety of supply and waste processing, and the safety of future reactors. Major actors are indicated: CEA, Areva, EDF, SEIV Alcen, Toshiba, Rolls Royce, and Comex. Some key data are indicated: expected lifetime, expected availability rate, cost. The projected site is Marcoule and fast breeder reactors operated or under construction in the world are indicated. The document also proposes an overview of the background and evolution of reactors of 4. generation

  14. On-site experimental dynamic analysis for evaluating the soil-structure interaction and the seismic behaviour of the Italian PEC fast reactor building

    International Nuclear Information System (INIS)

    Casirati, M.; Castoldi, A.; Panzeri, P.; Pezzoli, P.; Martelli, A.; Masoni, P.; Brancati, V.

    1988-01-01

    The paper describes the on-site dynamic tests carried out on the PEC fast reactor building, using various excitation methods (two eccentric back-rotating-mass mechanical vibrator, blasting in bore-hole, hydraulic actuators at the building foundations). It points out the purposes of the four tests campaigns performed at various construction stages and reports the main experimental results. These results concern both the design safety margins and the data for the validation of the three-dimensional numerical model of the reactor building, including soil-structure interaction phenomena. (author)

  15. Fast reactors: the industrial perspective

    International Nuclear Information System (INIS)

    Vaughan, R.D.

    1986-01-01

    Industrial participation in the development of the fast reactor is reviewed, from the construction of PFR at Dounreay to the initial steps towards collaboration in Europe. The optimum design of the fast reactor has changed considerably from the days when it was needed urgently to forestall a shortage of uranium to today when uranium is abundant and cheap. The evolution of the reactor design over this period is described. Collaboration in Europe is shown to be the only answer to high development costs and the search for a reactor which will compete with thermal reactors in today's environment. The partner countries in this collaboration are all motivated differently, and this is leading to some delays in concluding the necessary agreements. The objective on the industrial front is now to participate in the two or three demonstration fast reactors that will be built in Europe during the remainder of the century leading, it is hoped, to a competitive reactor design by the year 2000. (author)

  16. Experimental study of the large-scale axially heterogeneous liquid-metal fast breeder reactor at the fast critical assembly: Power distribution measurements and their analyses

    International Nuclear Information System (INIS)

    Iijima, S.; Obu, M.; Hayase, T.; Ohno, A.; Nemoto, T.; Okajima, S.

    1988-01-01

    Power distributions of the large-scale axially heterogeneous liquid-metal fast breeder reactor were studied by using the experiment results of fast critical assemblies XI, XII, and XIII and the results of their analyses. The power distributions were examined by the gamma-scanning method and fission rate measurements using /sup 239/Pu and /sup 238/U fission counters and the foil irradiation method. In addition to the measurements in the reference core, the power distributions were measured in the core with a control rod inserted and in a modified core where the shape of the internal blanket was determined by the radial boundary. The calculation was made by using JENDL-2 and the Japan Atomic Energy Research Institute's standard calculation system for fast reactor neutronics. The power flattening trend, caused by the decrease of the fast neutron flux, was observed in the axial and radial power distributions. The effect of the radial boundary shape of the internal blanket on the power distribution was determined in the core. The thickness of the internal blanket was reduced at its radial boundary. The influence of the internal blanket was observed in the power distributions in the core with a control rod inserted. The calculation predicted the neutron spectrum harder in the internal blanket. In the radial distributions of /sup 239/Pu fission rates, the space dependency of the calculated-to-experiment values was found at the active core close to the internal blanket

  17. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1979-10-01

    Various experiments being performed at the SNR reactor are described including: capture cross sections of various nuclei; fuel can failure; creep testing of welded joints; gas leakage through concrete/steel interfaces; testing of the test section of the four rod bundle for Laser Doppler Anemometry

  18. A review of the UK fast reactor programme, March 1979

    International Nuclear Information System (INIS)

    Smith, R.D.

    1979-01-01

    The Status report of the UK activities related to fast-breeder reactor activities includes the following: summary of the operating experience of the prototype Fast Reactor (PFR) during 1978; design studies of the commercial demonstration fast reactor (CDFR); design studies of later advanced LMFBR; engineering developments of high temperature sodium loop, steam generators and instrumentation; materials development; corrosion problems; sodium technology; fuel elements development; PFR fuel reprocessing; safety issues molten fuel-coolant interaction; core structure test; accident analysis; reactor performance studies; experimental reactor physics; fuel management and general neutronics calculation for CDFR; reactor instruments

  19. A review of the UK fast reactor programme, March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R D

    1979-07-01

    The Status report of the UK activities related to fast-breeder reactor activities includes the following: summary of the operating experience of the prototype Fast Reactor (PFR) during 1978; design studies of the commercial demonstration fast reactor (CDFR); design studies of later advanced LMFBR; engineering developments of high temperature sodium loop, steam generators and instrumentation; materials development; corrosion problems; sodium technology; fuel elements development; PFR fuel reprocessing; safety issues molten fuel-coolant interaction; core structure test; accident analysis; reactor performance studies; experimental reactor physics; fuel management and general neutronics calculation for CDFR; reactor instruments.

  20. Status of the DEBENE fast breeder reactor development, March 1979

    International Nuclear Information System (INIS)

    Daeunert, U.; Kessler, G.

    1979-01-01

    Status report of the Fast-breeder reactor development in Germany covers the following: description of the political situation in Federal republic of germany during 1978; international cooperation in the field of fast reactor technology development; operation description of the KNK-II fast core experimental power plant; status of construction of the SNR-300; results of the research and development programs concerned with fuel element, cladding, absorber rods and core structural materials development; sodium effects; neutron irradiation effects on SS properties; reactor physics related to experiments in fast critical assemblies; fast reactor safety issues; core disruption accidents; sodium boiling experiments, measuring methods developed; component tests

  1. Status of the DEBENE fast breeder reactor development, March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Daeunert, U; Kessler, G

    1979-07-01

    Status report of the Fast-breeder reactor development in Germany covers the following: description of the political situation in Federal republic of germany during 1978; international cooperation in the field of fast reactor technology development; operation description of the KNK-II fast core experimental power plant; status of construction of the SNR-300; results of the research and development programs concerned with fuel element, cladding, absorber rods and core structural materials development; sodium effects; neutron irradiation effects on SS properties; reactor physics related to experiments in fast critical assemblies; fast reactor safety issues; core disruption accidents; sodium boiling experiments, measuring methods developed; component tests.

  2. Fast Reactors and Nuclear Nonproliferation

    International Nuclear Information System (INIS)

    Avrorina, E.N.; Chebeskovb, A.N.

    2013-01-01

    Conclusion remarks: 1. Fast reactor start-up with U-Pu fuel: – dependent on thermal reactors, – no needs in U enrichment, – needs in SNF reprocessing, – Pu is a little suitable for NED, – practically impossible gun-type NED, – difficulties for implosion-type NED: necessary tests, advanced technologies, etc. – Pu in blankets is similar to WPu by isotopic composition, – Use of blanket for production isotopes (e.g. 233 U), – Combined reprocessing of SNF: altogether blanket and core, – Blanket elimination: decrease in Pu production – No pure Pu separation. 2. Fast reactor start-up with U fuel: - Needs in both U enrichment and SNF reprocessing, - Independent of thermal reactors, - Good Pu bred in the core let alone blankets, - NED of simple gun-type design, - Increase of needs in SWU, - Increased demands in U supply. 3. Fast reactors for export: - Uranium shortage, - To replace thermal reactors in future, - No blankets (depends on the country, though), - Fuel supply and SNF take back, - International centers for rendering services of NFC. Time has come to remove from FRs and their NFC the label unfairly identifying them as the most dangerous installations of nuclear power from the standpoint of being a proliferation problem

  3. The safety of the fast reactor

    International Nuclear Information System (INIS)

    Matthews, R.R.

    1977-01-01

    Verbatim of an address by R.R. Matthews, Chief Nuclear Health and Safety Officer, UK Central Electricity Generating Board given on January 15th 1977. The object of this address was to give some opinions on the safety issues of fast reactors as seen from an operational point of view. An outline of the basic responsibilities for nuclear safety is first given, and it is emphasized that the Central Electricity Generating Board has a statutory responsibility for the safe operation of its nuclear plant. The Nuclear Installations Act places absolute responsibility on the operator for ensuring that injury to persons and damage to property do not occur, and the new Health and Safety at Work Act does likewise. In addition the Board has a Nuclear Health and Safety Department that has to ensure that adequate provision for safety is made in the design, construction, and operation of nuclear plant, and safety at operational stations is monitored continuously by inspectors. In addition the requirements of the Nuclear Installations Inspectorate, laid down in the site licence conditions, must be satisfied. All these requirements are here discussed in the light of application to commercial fast reactors. It is considered that the hazards to fast reactor operating personnel are small and little different from those of other types of reactor, and in some respects the fast reactor has advantages, particularly in regard to the use of a Na coolant. The possibility of various types of accident is considered. Radioactive effluent discharge is also considered. The fast reactor as an international problem is discussed, including security matters. The extensive experience gained in operation of the experimental and prototype fast reactors at Dounreay is emphasized. (U.K.)

  4. Status of Fast Reactor Research and Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-01

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  5. Status of Fast Reactor Research and Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  6. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2012-01-01

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  7. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2013-01-01

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  8. The Integral Fast Reactor concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes the key features and potential advantages of the IFR concept, its technology development status, fuel cycle economics potential, and its future development path

  9. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Hudina, M.; Pelloni, S.; Sigg, B.; Stanculescu, A.

    1998-01-01

    The small Swiss research program on fast reactors serves to further understanding of the role of LMFR for energy production and to convert radioactive waste to more environmentally benign forms. These activities are on the one hand the contribution to the comparison of advanced nuclear systems and bring on the other to our physical and engineers understanding. (author)

  10. Review of fast reactor activities

    Energy Technology Data Exchange (ETDEWEB)

    Balz, W [Commission of the European Communities, Brussels (Belgium)

    1978-07-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle.

  11. Review of fast reactor activities

    International Nuclear Information System (INIS)

    Balz, W.

    1978-01-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle

  12. 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.)

  13. Fast reactor development programme in France

    Energy Technology Data Exchange (ETDEWEB)

    Le Rigoleur, C [Direction des Reacteurs Nucleaires, CEA Centre d` Etudes de Cadarache, Saint-Paul-lez-Durance (France)

    1998-04-01

    First the general situation regarding production of electricity in France is briefly described. Then in the field of Fast Reactors, the main events of 1996 are presented. At the end of February 1996, the PHENIX reactor was ready for operation. After review meetings, the Safety Authority has requested safety improvements and technical demonstrations, before it examines the possibility of authorizing a new start-up of PHENIX. The year 1996 was devoted to this work. In 1996, SUPERPHENIX was characterized by excellent operation throughout the year. The reactor was restarted at the end of 1995 after a number of minor incidents. The reactor power was increased by successive steps: 30% Pn up to February 6, followed by 50% Pn up to May then 60% up to October and 90% Pn during the last months. A programmed shutdown period occurred during May, June and mid-July 1996. The reactor has been shutdown at the end of 1996 for the decenial control of the steam generators. The status of the CAPRA project, aimed at demonstrating the feasibility of a fast reactor to burn plutonium at as high a rate as possible and the status of the European Fast Reactor are presented as well as their evolution. Finally the R and D in support of the operation of PHENIX and SUPERPHENIX, in support of the ````knowledge-acquisition```` programme, and CAPRA and EFR programmes is presented, as well as the present status of the stage 2 dismantling of the RAPSODIE experimental fast reactor. (author). 4 refs, figs, 2 tabs.

  14. Status of national programmes on fast reactors

    International Nuclear Information System (INIS)

    1994-04-01

    Based on the International Working Group on Fast reactors (IWGFR) members' request, the IAEA organized a special meeting on Fast Reactor Development and the Role of the IAEA in May 1993. The purpose of the meeting was to review and discuss the status and recent development, to present major changes in fast reactor programmes and to recommend future activities on fast reactors. The IWGFR took note that in some Member States large prototypes have been built or are under construction. However, some countries, due to their current budget constraints, have reduced the level of funding for research and development programmes on fast reactors. The IWGFR noted that in this situation the international exchange of information and cooperation on the development of fast reactors is highly desirable and stressed the importance of the IAEA's programme on fast reactors. These proceedings contain important and useful information on national programmes and new developments in sodium cooled fast reactors in Member States. Refs, figs and tabs

  15. Status of fast reactor activities in Brazil

    International Nuclear Information System (INIS)

    Menezes, Artur

    1996-01-01

    This text describes the present status of fast reactor activities in Brazil, emphasizing the strategies being used to preserve this reactor concept as a viable alternative for future electricity generation in the country. The program is mostly research-oriented and has the objective of establishing a consistent knowledge basis which can serve as a support for the transition to the activities more directly related to design, construction and operation of an experimental fast reactor. Due to the present economic difficulties, the program is still modest but it is gradually growing. A report which has been finalized in December, 1995 and submitted to the authorities indicates the existence of the grounds for enlarging and consolidating the program. (author)

  16. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  17. Status of liquid metal cooled fast reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants Refs, figs, tabs

  18. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

  19. Fast reactor core monitoring device

    International Nuclear Information System (INIS)

    Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

    1982-01-01

    Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

  20. Simulating the Behaviour of the Fast Reactor Joyo (Draft)

    International Nuclear Information System (INIS)

    Juutilainen, Pauli

    2008-01-01

    Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO 2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)

  1. Fast reactor programme

    International Nuclear Information System (INIS)

    Slagter, W.; Roodbergen, H.A.

    1975-07-01

    A numerical method has been applied to determine three-dimensional temperature distributions in small regions of hexagonal fuel rod assemblies with liquid metal cooling. The temperature calculations, based on the finite element method, centered on heat transport from the fuel rod surface into the adjacent subchannel and on heat transport between adjacent subchannels. The study is restricted to analysis of steady-state heat transfer in fuel rod assemblies of smooth parallel rods for the condition of prescribed heat flux and slug flow. The computed temperature profiles are expressed in terms of Nusselt or Stanton numbers and amplitudes of circumferential cladding temperature variations as a function of axial distances. In simple geometries like the circular duct, the method is applied to calculate the temperature distribution and the axial variation of the Nusselt number for the condition of laminar flow and uniform heat flux. The flow is hydrodynamically developed. Good agreement between numerical and theoretical solutions confirms the validity of the finite element method. The results obtained for fuel rod assemblies are compared with available experimental and theoretical predictions. It is found that the contribution of turbulence to intersubchannel heat transport in sodium equals that due to conduction. The novel feature of the method applied to convective heat transfer problems is the computation of detailed three-dimensional coolant temperature fields in large regions of fuel rod assemblies

  2. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1980-10-01

    Results of the estimate of the contribution of (n,p) and (n,α) reactions to the capture cross sections of individual and lumped fission-products are discussed; the progress in evaluating neutron cross sections of corrosion products is reported and some recent results of the study on pre-equilibrium models are given. The last canning failure experiment has been performed, and some results of the post-irradiation examination of two loss-of-cooling experiments are elucidated; the progress in constructing and testing the equipment for the transient overpower experiments is reported. The proceedings in the thermochemical investigations on different uranium compounds are described. Creep strength data of three heats and two types of welded joints of DIN 1.4948 parent metal in irradiated and reference condition are intercompared, and the effect of the strain rate on the low-cycle fatigue behaviour is shown; a MONA-rig has been developed for the investigation of fluence effects on stainless steel at high temperatures; a preliminary experiment has revealed the minimum specimen dimensions for fracture toughness testing to obtain valid test results according to ASTM directives. Attempts have been made to produce sodium smoke aerosols and to characterize them by determining their shape factors; a system has been built to measure accurately the leak flows through capillary leaks. Methods have been developed to determine accurately the wall positions in channels of complicated geometries for the application of LDA; a brief review is given of all the boiling experiments performed in bundle I (68,5% blockage) and bundle II (34,5% blockage); the objectives of a profound analysis of the experimental data of all these boiling experiments are given

  3. Intermediate and fast neutron absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-10-01

    The experimental fuel channel EFC is created as one of the fast neutron fields at the RB reactor. The intermediate and fast neutron spectra in EFC are measured by activation technique. The intermediate and fast neutron absorbed doses are computed on the basis of these experimental results. At the end the obtained doses are compared. (author)

  4. Fuel management codes for fast reactors

    International Nuclear Information System (INIS)

    Sicard, B.; Coulon, P.; Mougniot, J.C.; Gouriou, A.; Pontier, M.; Skok, J.; Carnoy, M.; Martin, J.

    The CAPHE code is used for managing and following up fuel subassemblies in the Phenix fast neutron reactor; the principal experimental results obtained since this reactor was commissioned are analyzed with this code. They are mainly concerned with following up fuel subassembly powers and core reactivity variations observed up to the beginning of the fifth Phenix working cycle (3/75). Characteristics of Phenix irradiated fuel subassemblies calculated by the CAPHE code are detailed as at April 1, 1975 (burn-up steel damage)

  5. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  6. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: a pool-type primary system, and advanced ternary alloy metallic fuel, and an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  7. The 'SURA' fast reactor program

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The Commissariat a l'Energie Atomique's SURA program on fast reactor safety consists of two specific testing programs on fastbreeder reactor safety: the Cabri and Scarabee programs. Both Cabri and Scarabee are examples of multinational research collaboration. The CEA and the Karlsruhe Nuclear Research Center are each covering half of the construction costs. Britain, the US and Japan are also due to participate in these experiments. The aim of the programs is to examine the behaviour of fuel in sodium cooled fast reactors. The Cabri program consists of setting off a reactivity accident in a power reactor core which is cooled with liquid sodium, such an accident occurring after a sharp increase in reactivity or as a result of the pump suddenly breaking down without there at the same time being any fall in the control rods. In 1967 the Commissariat a l'Energie Atomique started its Scarabee research program which is trying to analyse the sort of things that can go wrong with fuel cooling systems and what the consequences can be [fr

  8. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  9. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Blevins, J.D.; Stasko, R.R.

    1989-09-01

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  10. Orphee reactor experimental equipment

    International Nuclear Information System (INIS)

    1987-01-01

    Experimental equipment around the ORPHEE reactor is presented. The neutron source; and the spectrometers and sample environment (inelastic and quasi-elastic scattering, elastic scattering, spread scattering, small angle scattering) are described. An experiment proposal and reports guide is supplied [fr

  11. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Bourgeois, M.; Le Bouhellec, J.; Eymery, R.; Viala, M.

    1984-08-01

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  12. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. Recovery of MARICO-2 sample part

    International Nuclear Information System (INIS)

    Ashida, Takashi; Ito, Hideaki

    2015-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. The following items are introduced here: (1) summary of restoration work and overall process of restoration work, (2) recovery operation of MARICO-2 sample part, (3) exchange of the upper core structure that was conducted this year, and (4) results of recovery of MARIKO-2 sample part. (A.O.)

  13. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 1. MARICO-2 subassembly retrieval work

    International Nuclear Information System (INIS)

    Naito, Hiroyuki; Ashida, Takashi; Ito, Hideaki

    2014-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. This paper introduces the progress of restoration work and the future work plan, with a focus on the outline of overall restoration work, the method / problems / measures for MARICO-2 sample part recovery operations, and fabrication of sample part recovery device. (A.O.)

  14. Advances by the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs

  15. Electrochemistry in fast reactor technology

    International Nuclear Information System (INIS)

    Mathews, C.K.

    1987-01-01

    Electrochemistry plays a significant role in the production, characterisation or behaviour of the fuel, the coolant and structural materials used in fast reactor systems. The role of electrochemistry in sodium production, in the fuel cycle, in the development of electrochemical meters used for the on-line monitoring of the various impurities at sub ppm levels and in the recovery of plutonium and uranium are discussed. The advantage of voltammmetric techniques in the analysis of impurities and the application of electrochemical meters have been investigated. (author). 5 figs., 15 refs

  16. Fast reactor fuel reprocessing. An Indian perspective

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2005-01-01

    The Department of Atomic Energy (DAE) envisioned the introduction of Plutonium fuelled fast reactors as the intermediate stage, between Pressurized Heavy Water Reactors and Thorium-Uranium-233 based reactors for the Indian Nuclear Power Programme. This necessitated the closing of the fast reactor fuel cycle with Plutonium rich fuel. Aiming to develop a Fast Reactor Fuel Reprocessing (FRFR) technology with low out of pile inventory, the DAE, with over four decades of operating experience in Thermal Reactor Fuel Reprocessing (TRFR), had set up at the India Gandhi Center for Atomic Research (IGCAR), Kalpakkam, R and D facilities for fast reactor fuel reprocessing. After two decades of R and D in all the facets, a Pilot Plant for demonstrating FRFR had been set up for reprocessing the FBTR (Fast Breeder Test Reactor) spent mixed carbide fuel. Recently in this plant, mixed carbide fuel with 100 GWd/t burnup fuel with short cooling period had been successfully reprocessed for the first time in the world. All the challenging problems encountered had been successfully overcome. This experience helped in fine tuning the designs of various equipments and processes for the future plants which are under construction and design, namely, the DFRP (Demonstration Fast reactor fuel Reprocessing Plant) and the FRP (Fast reactor fuel Reprocessing Plant). In this paper, a comprehensive review of the experiences in reprocessing the fast reactor fuel of different burnup is presented. Also a brief account of the various developmental activities and strategies for the DFRP and FRP are given. (author)

  17. The integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1990-01-01

    The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management

  18. A review of the UK fast reactor programme. March 1977

    International Nuclear Information System (INIS)

    Smith, R.D.

    1977-01-01

    This paper reports on the Fast Reactor Programme of United Kingdom. These are the main lines: Dounreay Fast Reactor; Prototype Fast Reactor; Commercial Fast Reactor; engineering development; materials development; chemical engineering/sodium technology; fast reactor fuel; fuel cycle; safety; reactor performance study

  19. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Kapoor, R.P.; Srinivasan, G.; Ellappan, T.R.; Ramalingam, P.V.; Vasudevan, A.T.; Iyer, M.A.K.; Lee, S.M.; Bhoje, S.B.

    2000-01-01

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  20. Determination of homogenization factors in the simulation in the azymuthal flux unit distribution in the control ring of a experimental fast reactor

    International Nuclear Information System (INIS)

    Jachic, J.

    1987-01-01

    The azimuthal neutronic flux distribution in the control ring region for a low power fast reactor is simulated using a plate and rectangular smash models for one dimensional calculations under periodic boundary conditions in the frontier. (E.G.) [pt

  1. Actinide behavior in the integral fast reactor

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

  2. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  3. History of fast reactor development in U.S.A.-I

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Sasao, Nobuyki

    2007-01-01

    History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

  4. Current status of restoration work for obstacle and supper core structure in reactor vessel of experimental fast reactor 'JOYO'. 3. Sodium purification operation after MARICO recovery and UCS exchange work

    International Nuclear Information System (INIS)

    Shimizu, Shunji; Izawa, Osamu; Ishizaki, Kazuhiko; Takeishi, Tsuyoshi; Oowada, Ryohei; Yoshihara, Shizuya; Michino, Masanobu

    2015-01-01

    At fast-breeder reactor 'Joyo', in order to restore the partial inhibition of the rotating plug fuel exchange function due to interference with 'experimental apparatus with instrumentation lines (MARICO-2)', which occurred in May 2007, a recovery work was performed. The replacement work of the upper core structure and the recovery of sample part of the experimental apparatus with instrumentation lines were carried out under conditions where the primary system sodium was drained and the liquid level of reactor vessel was lowered. During the pulling-up work of upper core structure, an increase in nitrogen and hydrogen concentrations in the reactor vessel cover gas (argon) was confirmed through the measurement of the primary system gas chromatograph. This was due to the intrusion of air caused by the opening of the cover gas boundary. Since entrained oxygen reacted with sodium in the reactor, the purity of sodium was reduced. When this sodium is purified according to common method, the sodium with decreased purity defuses through the entire primary cooling system, causing various adverse effects. A safe and reliable procedure to purify sodium while preventing the adverse effects was examined and practiced. (A.O.)

  5. Holography for fast reactor inspection

    International Nuclear Information System (INIS)

    Tozer, B.A.

    1980-01-01

    Holography, an optical process whereby an image of the original subject can be reconstructed in three dimensions, is being developed for use as an optical inspection tool. With a potential information storage density of 10 16 bits/m 2 , the ability to reconstruct in 3 dimensions, a depth of field of up to 8 metres, extremely wide angle of view, and potentially diffraction limited resolution, holography should be invaluable for the optical recording of fast reactors during construction, and the inspection of optically accessible regions during operation, or maintenance down-times. The photographic emulsions used for high resolution holography are fine-grained and fog only very slowly when subjected to γ-radiation, so that inspection of highly radio-active regions and components can be effected satisfactorily. Some of the practical limitations affecting holography are described and ways of overcoming them discussed. Some preliminary results are presented. (author)

  6. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  7. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  8. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  9. Research activities on fast reactors in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Dones, R.; Hudina, M.; Pelloni, S.

    1996-01-01

    The current domestic Swiss electricity supply is primarily based on hydro power (approximately 61%) and nuclear power (about 37%). The contribution of fossil systems is, consequently, minimal (the remaining 2%). In addition, long-term (but limited in time) contracts exist, securing imports of electricity of nuclear origin from France. During the last two years, the electricity consumption has been almost stagnant, although the 80s recorded an average annual increase rate of 2.7%. The future development of the electricity demand is a complex function of several factors with possibly competing effects, like increased efficiency of applications, changes in the industrial structure of the country, increase of population, further automation of industrial processes and services. Due to decommissioning of the currently operating nuclear power plants and expiration of long-term electricity import contracts there will eventually open a gap between the postulated electricity demand and the base supply. The assumed projected demand cases, high and low, as well as the secured yearly electric energy supply are shown. The physics aspects of plutonium burning fast reactor configurations are described including first results of the CIRANO experimental program. Swiss research related to residual heat removal in fast breeder reactors is presented. It consists of experimental ana analytic investigations on the mixing between two horizontal fluid layers of different velocities and temperatures. Development of suitable computer codes for mixing layer calculation are aimed to accurately predict the flow and temperature distribution in the pools. A satisfactory codes validation based on experimental data should be done

  10. The UK commercial demonstration fast reactor design

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1987-01-01

    The paper on the UK Commercial Demonstration Fast Reactor design was presented to the seminar on 'European Commercial Fast Reactor Programme, London 1987. The design is discussed under the topic headings:- primary circuit, intermediate heat exchangers and pumps, fuel and core, refuelling, steam generators, and nuclear island layout. (U.K.)

  11. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    Zouain, D.M.

    1982-06-01

    The fuel cycle for fast reactors, is analysed, regarding the technical aspects of the developing of the reprocessing stages and the fuel fabrication. The environmental impact of LMFBRs and the waste management of this cycle are studied. The economic aspects of the fuel cycle, are studied too. Some coments about the Brazilian fast reactors programs are done. (E.G.) [pt

  12. 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

  13. Euratom contributions in Fast Reactor research programmes

    International Nuclear Information System (INIS)

    Fanghänel, Th.; Somers, J.

    2013-01-01

    The Sustainable Nuclear Initiative: • demonstrate long-term sustainability of nuclear energy; • demonstration reactors of Gen IV: •more efficient use of resources; • closed fuel cycle; • reduced proliferation risks; • enhanced safety features. • Systems pursued in Europe: • Sodium-cooled fast reactor SFR; • Lead-cooled fast reactor LFR; • Gas-cooled fast reactor GFR. Sustainable Nuclear Energy Technology Platform SNE-TP promotes research, development and demonstration of the nuclear fission technologies necessary to achieve the SET-Plan goals

  14. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  15. Economic Issues of Fast Reactor in China

    International Nuclear Information System (INIS)

    Yang Hongyi

    2013-01-01

    Conclusions: 1. More and more fast reactors could be appearing in the world currently and near future. 2. China gets little experience and practice about the economics issues of sodium cooled fast reactors. 3. The economic issues become more and more important for the deplot of fast reactors. Suggestions: 1. An authoritative economic evaluation solution for fast reactor and related fuel cycles facilities is necessary. The solution may be developed by the interested country in order to share the few data, experience and methodology. 2. A new initiative to help to share the economic information for fast reactor and related fuel cycle facilities is necessary. A meeting like TM-44899 organized by the IAEA is very beneficial for this topic and hopefully will continue

  16. Status of national programmes on fast reactors 1997/98. 31. annual meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    1998-01-01

    The objective of the meeting was to co-ordinate the exchange of information on the status of fast reactor development and operational experience, including experience with experimental types of reactor; to consider meeting arrangements for 1998 and 1999; and to review the IAEA co-ordinated research activities in the field of fast reactor, as well as co-ordination of the International Working Group on Fast Reactors activities with other organizations

  17. Status of fast reactor activities in Russia

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Yu.M.; Zverev, K.V.

    1998-01-01

    This paper outlines state-of-the-art of the Russian nuclear power as of 1997 and its prospects for the nearest future. Results of the BR-10, BOR-60 and BN-600 reactors operation are described, as well as activity of the Russian institutions on scientific and technological support of the BN-350 reactor. Analysis of current status of the BN-800 reactor South-Urals NPP and Beloyarskaya NPP designs is given in brief, as well as prospects of their construction and possible ways of fast reactor technology improvement. Studies on fast reactors now under way in Russia are described. (author)

  18. Fast reactor fuel reprocessing in the UK

    International Nuclear Information System (INIS)

    Allardice, R.H.; Williams, J.; Buck, C.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the U.K. since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium based fast reactor system and the importance of establishing at an early stage fast reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high burn-up thermal reactor oxide fuel. In consequence, the U.K. has decided to reprocess irradiated fuel from the 250 MW(E) Prototype Fast Reactor as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small scale fully active demonstration plant have been carried out over the past 5 years and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant a parallel development programme has been initiated to provide the basis for the design of a large scale fast reactor fuel reprocessing plant to come into operation in the late 1980s to support the projected U.K. fast reactor installation programme. The paper identifies the important differences between fast reactor and thermal reactor fuel reprocessing technologies and describes some of the development work carried out in these areas for the small scale P.F.R. fuel reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast reactor fuel reprocessing plant is outlined and the current design philosophy is discussed

  19. General remarks on fast neutron reactor physics

    International Nuclear Information System (INIS)

    Barre, J.Y.

    1980-01-01

    The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

  20. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  1. 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.

  2. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  3. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

    2012-01-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  4. Design characteristics of research zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Nikolic, D.; Antic, D.; Zavaljevski, N.; Popovic, D.

    1990-01-01

    LASTA is a flexible zero power reactor with uranium and plutonium fuel designed for research in the neutron physics and in the fast reactor physics. Safety considerations and experimental flexibility led to the choice of a fixed vertical assembly with two safety blocks as the main safety elements, so that safety devices would be operated by gravity. The neutron and reactor physics, the control and safety philosophy adopted in our design, are described in this paper. Developed computer programs are presented. (author)

  5. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  6. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements

  7. Fission energy: The integral fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

  8. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  9. Safeguards challenges of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Ko, H. S.

    2010-01-01

    Although the safeguards system of Sodium Fast Reactor (SFR) seems similar to that of Light Water Reactor (LWR), it was raised safeguards challenges of SFR that resulted from the visual opacity of liquid sodium, chemical reactivity of sodium and other characteristics of fast reactor. As it is the basic concept stage of the safeguards of SFR in Korea, this study tried to analyze the latest similar study of safeguards issues of the Fast Breeder Reactor (FBR) at Joyo and Monju in Japan. For this reason, this study is to introduce some potential safeguards challenges of Fast Breeder Reactor. With this analysis, future study could be to address the safeguards challenges of SFR in Korea

  10. Power from plutonium: fast reactor fuel

    International Nuclear Information System (INIS)

    Bishop, J.F.W.

    1981-01-01

    Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

  11. Development of physical conceptions of fast reactors

    International Nuclear Information System (INIS)

    Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

    2013-01-01

    • Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

  12. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of 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 the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  13. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.; Bando, S.

    1981-03-01

    The fast breeder reactor development project in Japan made progress in the past year, and will be continued in the next fiscal 1981. The scale of efforts both in budget and personnel will be similar to those in fiscal 1980. The budget for R and D works and for the construction of the fast breeder prototype reactor ''Monju'' will be approximately 20 billion yen and 27 billion yen, respectively, excluding the wage of the personnel concerned. The number of the technical personnel currently engaging in fast breeder reactor development in the Power Reactor and Nuclear Fuel Development Corp. is about 530. As for the experimental fast reactor ''Joyo'', three operational cycles at 75 MWt have been completed in August, 1980, and the fourth cycle has started in March, 1981. As for the prototype reactor ''Monju'', progress was made toward the construction, and the environmental impact statement on the reactor was approved by the authorities concerned. The studies on the preliminary design of large LMFBRs have been made by the PNC and also by power companies. The design study carried out by the PNC is concerned with a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of the commissioning of ''Monju''. The highlights and topics in the development activities for fast breeder reactors in the past twelve months are summarized in this report. (Kako, I.)

  14. Some questions and answers concerning fast reactors

    International Nuclear Information System (INIS)

    Marshall, W.

    1980-01-01

    The theme of the lecture is the place of the fast reactor in an evolving nuclear programme. The whole question of plutonium is first considered, ie its method of production and the ways in which it can be used in the fast reactor fuel cycle. Whether fast reactors are necessary is then discussed. Their safety is examined with particular attention to those design features which are most criticised ie high volumetric power density of the core, and the use of liquid sodium as coolant. Attention is then paid to environmental and safeguard aspects. (U.K.)

  15. Fast breeder reactors--lecture 4

    International Nuclear Information System (INIS)

    Marshall, W.; Davies, L.M.

    1986-01-01

    This paper discusses the economics of fast breeder reactors. An algebraic background is presented which represents the various views expressed by different nations regarding the cost of fast breeder reactors and their associated fuel cycle services, the timescale by which they might be available, and the simultaneous variations in the price of uranium. Actual presentations made by individual countries in recent discussions serve to verify the general nature of this present discussion. It is assumed that if nuclear power is to make a long term contribution to the needs of the world, the introduction of fast breeder reactors is both essential and necessary

  16. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  17. Philosophy of safety evaluation on fast breeder reactor

    International Nuclear Information System (INIS)

    1981-01-01

    This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

  18. Sodium fires at fast reactors: RF status report

    International Nuclear Information System (INIS)

    Bagdasarov, Yu.E.; Buksha, Yu.K.; Drobyshev, A.V.; Zybin, V.A.; Ivanenko, V.N.; Kardash, D.Yu.; Kulikov, E.V.; Yagodkin, I.V.

    1996-01-01

    Scientific and engineering studies carried out in Russian Federation since 1992 up to 1996 in the sodium fire area and their main results are described. A review of activities on modification of the computer codes BOX and AERO developed at IPPE for calculating sodium fire consequences is given. Results of analysis of possible accidental situations at currently designed BN-800 reactor NPP with the use of these codes are presented. Sodium leaks occurring at our domestic fast reactors are briefly analyzed. Experimental work performed are described. Results of comparative analysis of common-cause and sodium fire hazards for fast reactor NPP are presented. (author)

  19. A Comparison of Theoretical and Experimental Values of the Activation Doppler Effect in Some Fast Reactor Spectra

    International Nuclear Information System (INIS)

    Haeggblom, H.; Tiren, L.I.

    1968-08-01

    Results of activation Doppler measurements on the U 238 (n,γ) and U 235 (n, fission) reactions in the FR0 and MSCA fast critical assemblies have been compared with theoretical values. The study covers neutron spectra with median fission energies from 50 to 240 keV. The calculated Doppler effect in U 238 in the FR0 cores is 20 - 35 % lower than the measured values. The sensitivity of the theoretical result with regard to changes in cross sections and neutron spectrum has been studied. The theoretical value for U 235 (FR0 core 5) is 4 times higher than the measured one. The report includes a brief description of the DORIX-2 method of calculating effective resonance cross sections appropriate to activation Doppler measurements. References to the cross section data used for the comparisons are also given

  20. A Comparison of Theoretical and Experimental Values of the Activation Doppler Effect in Some Fast Reactor Spectra

    Energy Technology Data Exchange (ETDEWEB)

    Haeggblom, H; Tiren, L I

    1968-08-15

    Results of activation Doppler measurements on the U{sup 238} (n,{gamma}) and U{sup 235} (n, fission) reactions in the FR0 and MSCA fast critical assemblies have been compared with theoretical values. The study covers neutron spectra with median fission energies from 50 to 240 keV. The calculated Doppler effect in U{sup 238} in the FR0 cores is 20 - 35 % lower than the measured values. The sensitivity of the theoretical result with regard to changes in cross sections and neutron spectrum has been studied. The theoretical value for U{sup 235} (FR0 core 5) is 4 times higher than the measured one. The report includes a brief description of the DORIX-2 method of calculating effective resonance cross sections appropriate to activation Doppler measurements. References to the cross section data used for the comparisons are also given.

  1. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  2. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    2009-11-01

    Acknowledging the importance of passing on knowledge and experience, as well mentoring the next generation of scientists and engineers, and in response to expressed needs by Member States, the IAEA has undertaken concrete steps towards the implementation of a fast reactor data retrieval and knowledge preservation initiative. Decommissioning of fast reactors and other sodium bearing facilities is a domain in which considerable experience has been accumulated. Within the framework and drawing on the wide expertise of the Technical Working Group on Fast Reactors (TWG-FR), the IAEA has initiated activities aiming at preserving the feedback (lessons learned) from this experience and condensing those to technical recommendations on fast reactor design features that would ease their decommissioning. Following a recommendation by the TWG-FR, the IAEA had convened a topical Technical Meeting (TM) on 'Operational and Decommissioning Experience with Fast Reactors', hosted by CEA, Centre d'Etudes de Cadarache, France, from 11 to 15 March 2002 (IAEA-TECDOC- 1405). The participants in that TM exchanged detailed technical information on fast reactor operation and decommissioning experience with various sodium cooled fast reactors, and, in particular, reviewed the status of the various decommissioning programmes. The TM concluded that the decommissioning of fast reactors to reach safe enclosure presented no major difficulties, and that this had been accomplished mainly through judicious adaptation of processes and procedures implemented during the reactor operation phase, and the development of safe sodium waste treatment processes. However, the TM also concluded that, on the path to achieving total dismantling, challenges remain with regard to the decommissioning of components after sodium draining, and suggested that a follow-on TM be convened, that would provide a forum for in-depth scientific and technical exchange on this topic. This publication constitutes the Proceedings of

  3. The energy gap and the fast reactor

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The background to the development of fast reactors is summarized. In Britain, the results of the many experiments performed, the operation of the Dounreay Fast Reactor for the past 18 years and the first year's operation of the larger Prototype Fast Reactor have all been very encouraging, in that they demonstrated that the performance corresponded well with predictions, breeding is possible, and the system is exceptionally stable in operation. The next step in fast reactor engineering is to build a full-scale fast reactor power station. There would seem to be little reason to expect more trouble than could reasonably be expected in constructing any large project of this general nature. However, from an engineering point of view continuity of experience is required. If a decision to build a commercial fast reactor were taken today there would be a 14-year gap between strating this and the start of the Prototype Fast Reactor. This is already much too long. From an environmental standpoint we have to demonstrate that we can manufacture and reprocess fast reacctor fuel for a substantial programme in a way that does not lead to pollution of the environment, and that plutonium-containing fuel can be transported in the quantities required in safety and in a way that does not attract terrorists or require a private army to ensure its security. Finally, we have to find a way to allow many countries to obtain the energy they need from fast reactors, without leading to the proliferation of nuclear weapons or weapons capability. (author)

  4. Optimal reactor strategy for commercializing fast breeder reactors

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

    In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

  5. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  6. Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  7. The fast reactor and energy supply

    International Nuclear Information System (INIS)

    1979-01-01

    The progress made with fast reactor development in many countries is summarised showing that the aim is to provide to the nation concerned an ability to instal fast reactor power stations at the end of this century or early in the next one. Accepting the importance of fast reactors as a potential independent source of energy, problems concerning economics, industrial capability, technical factors, public acceptibility and in particular plutonium management, are discussed. It is concluded that although fast reactors have reached a comparatively advanced stage of development, a number of factors make it likely that their introduction for electricity generation will be a gradual process. Nevertheless it is necessary to complete demonstration and development phases in good time. (U.K.)

  8. Design codes for fast reactor steam generators

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1978-01-01

    The paper reviews the design methods and design criteria which are available for fast reactor structures, and discusses the materials data which are required to demonstrate the integrity of the plant components. (author)

  9. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  10. New experimental space for irradiating samples by RA reactor fast neutron flux at temperatures up to 100 deg C; Novi eksperimentalni prostori namenjeni ozracivanju uzoraka u fluksu brzih neutrona na temperaturama do 100{sup 0} C na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Novakovic, M; Zecevic, V [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1961-07-01

    The objective of this paper is to present adaptation of the RA reactor which would enable samples irradiation by fast neutrons and describe new experimental possibilities. New experimental space was achieved using hollow fuel elements which have been reconstructed to enable placement of irradiation capsules inside the tube. This paper includes thermal analysis and describes problems related to operation, safety and radiation protection issues which arise from using reconstructed fuel elements.

  11. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    International Nuclear Information System (INIS)

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  12. Review of fast reactor activities in India

    Energy Technology Data Exchange (ETDEWEB)

    Paranjpe, S R [Reactor Research Centre, Kalpakkam, Tamil Nadu (India)

    1981-05-01

    It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties.

  13. Fast reactors - Dounreay and the future

    International Nuclear Information System (INIS)

    Jordan, G.

    1988-01-01

    In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

  14. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1981-01-01

    It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties

  15. Advanced liquid metal fast breeder reactor designs

    International Nuclear Information System (INIS)

    Sayles, C.W.

    1978-01-01

    Fast Breeder reactor power plants in the 1000-1200 MW(e) range are being built overseas and are being designed in this country. While these reactors have many characteristics in common, a variety of different approaches have been adopted for some of the major features. Some of those alternatives are discussed

  16. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  17. A glossary of terms for fast reactors

    International Nuclear Information System (INIS)

    Wheeler, R.C.

    1979-04-01

    The glossary aims to provide definitions of technical terms likely to be used in a fast reactor enquiry and to encourage the use of the same set of consistent terms in any documents intended for such an inquiry. In some cases definitions are formulated in the limited context of LMFBRS rather than applying to all types of reactors. A brief guide is presented to the different reactor types. (author)

  18. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    Vautrey, L.; Zaleski, C.P.

    1962-01-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [fr

  19. Review of Fast Reactor Activities, March 1980

    International Nuclear Information System (INIS)

    Balz, W.

    1980-01-01

    As in previous years, a short outline of the major achievements made since the last IWGFR meeting is given in the following. On 18 February 1980 the Council of Ministers has approved a resolution in which they recognise the strategic importance of fast breeder reactors and the need to continue the efforts towards maintaining an effective fast breeder option in the Member States

  20. Integral measurement of fission products capture in fast breeder reactors

    International Nuclear Information System (INIS)

    Martin Deidier, Loick.

    1979-12-01

    For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

  1. 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.)

  2. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    that the cladding strength sufficient to withstand stress accounting for decreased thickness by the ACCI zone. (5) The wet wash and storage method was selected for disposing of the spent sodium bonded control rods, based upon experimental results at the JOYO facilities. The effects from storing sodium bonded control rods in wet storage were evaluated. The results indicated that these effect would not pose a safety problem. (author)

  3. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

    The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

  4. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-01-01

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  5. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  6. Calculation of the neutron parameters of fast thermal reactor

    International Nuclear Information System (INIS)

    Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.

    1975-01-01

    The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)

  7. Strategies for minority actinides transmutation in fast reactors

    International Nuclear Information System (INIS)

    Perez-Martin, S.; Martin-Fuertes, F.; Alvarez-Velarde, F.

    2010-01-01

    Presentation of the strategies that can be followed in fast reactors designed for the fourth generation to reduce the inventory of minority actinides generated in current light water reactors, as the actinides generation in fast reactor.

  8. A review of fast reactor programme in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Masuno, Y [Experimental Fast Reactor Division, O-arai Engineering Center, PNC (Japan); Bando, S [Project Planning and Management Division, PNC, Minato-ku, Tokyo (Japan)

    1981-05-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report.

  9. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    Masuno, Y.; Bando, S.

    1981-01-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

  10. Actinides burnup in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

  11. A worldwide survey of fast breeder reactors

    International Nuclear Information System (INIS)

    Hennies, H.H.

    1986-01-01

    While the completion of the SNR 300 was accompanied by manifold discussions on questions relevant to safety and energy policies in the Federal Republic of Germany and as a result considerable scheduling delays and exceeding of budgets were recorded, breeder reactor technology has been progressing worldwide. The transition from the development phase with small trial reactors to the construction and operation of large performance reactors was completed systematically, in particular in France and the Soviet Union. Even though the uranium supply situation does not make a short-term and comprehensive employment of fast breeder reactors essential, technology has meanwhile been advanced to such a level and extensive operating experience is on hand to enable the construction and safe operation of fast breeder reactors. A positive answer has long been found to the question of the realization of a breeding rate to guarantee the breeding effect. There remain now the endeavors to achieve a reduction in investment and fuel cycle costs. (orig.) [de

  12. Nuclear Burning Wave Modular Fast Reactor Concept

    International Nuclear Information System (INIS)

    Kodochigov, N.G.; Sukharev, Yu.P.

    2014-01-01

    The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

  13. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    , mainly pressurized heavy water reactors (PHWRs) to .... plug housing 12 absorber rod drive mechanisms is supported on ... state-of-art erection equipments and construction methodologies and .... This decision is taken after.

  14. Fast reactors and problems in their development. Chapter 6

    International Nuclear Information System (INIS)

    Dombey, N.

    1980-01-01

    The main differences between fast reactors, in particular the liquid-metal fast breeder reactor (LMFBR), and thermal reactors are discussed. The view is taken, based on the intrinsic physics of the systems, that fast reactors should be considered as a different genus from thermal reactors. Some conclusions are drawn for fast reactor development generally and for the British programme in particular. Physics, economics and safety aspects are covered. (U.K.)

  15. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1996-01-01

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

  16. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  17. Status of the fast breeder reactor technology in China

    International Nuclear Information System (INIS)

    Xu Mi

    1990-01-01

    According to the Chinese long-term energy strategy the FBR development is strongly supported. In the near term nuclear programme it is intended to build the experimental First Fast Reactor (FFR) in the year 2000. Design work is in progress. (author). 1 ref., 6 figs, 8 tabs

  18. Identification of fast power reactivity effect in nuclear power reactor

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  20. Fast-acting nuclear reactor control device

    International Nuclear Information System (INIS)

    Kotlyar, O.M.; West, P.B.

    1993-01-01

    A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

  1. Investigation of molten salt fast reactor

    International Nuclear Information System (INIS)

    Kubota, Kenichi; Konomura, Mamoru

    2002-01-01

    On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

  2. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    Parcy, J.P.

    1982-09-01

    A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented [fr

  3. Current status of fast reactor physics

    International Nuclear Information System (INIS)

    Hummel, H.H.

    1979-01-01

    The subject of calculation of reactivity coefficients for fast reactors is developed, starting with a discussion of the status of relevant nuclear data and proceeding to the subjects of group cross section generation and of methods of obtaining reactivity coefficients from group cross sections. Reactivity coefficients measured in critical experiments are compared with calculated values. Dependence of reactivity coefficients on reactor design is discussed. Finally, results of the recent international comparison of calculated reactivity coefficients are presented

  4. 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

  5. Slow clean-up for fast reactor

    Science.gov (United States)

    Banks, Michael

    2008-05-01

    The year 2300 is so distant that one may be forgiven for thinking of it only in terms of science fiction. But this is the year that workers at the Dounreay power station in Northern Scotland - the UK's only centre for research into "fast" nuclear reactors - term as the "end point" by which time the site will be completely clear of radioactive material. More than 180 facilities - including the iconic dome that housed the Dounreay Fast Reactor (DFR) - were built at at the site since it opened in 1959, with almost 50 having been used to handle radioactive material.

  6. Discharges from a fast reactor reprocessing plant

    International Nuclear Information System (INIS)

    Barnes, D.S.

    1987-01-01

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

  7. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Smith, R.D.

    1982-01-01

    A review of the United Kingdom Fast Reactor Programme is introduced. Operational experience with the Prototype Fast Reactor (PFR) is briefly summarized. The design concept of the Commercial Demonstration Fast Reactor (CDFR) is given in some detail. The emphasis is on materials development, chemical engineering/sodium technology, fuel reprocessing and fuel cycle, engineering component development and reactor safety

  8. Universal Fast Breeder Reactor Subassembly Counter manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

    1984-08-01

    A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

  9. The behaviour of materials in fast reactors

    International Nuclear Information System (INIS)

    Matthews, J.R.

    1977-01-01

    Fast neutron damage in fast reactors can limit the life of structural components through the growth voids. The main features of the current theory of point defect production and condensation are surveyed. The role of metallurgical structures and radiation produced extended defects is outlined and used to demonstrate the development of volume swelling and radiation hardening. Mechanisms of radiation creep are described in the context of the preceding treatment of point defect behaviour. Finally, future trends in the field are briefly explored. (author)

  10. Universal Fast Breeder Reactor Subassembly Counter manual

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

    1984-08-01

    A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications.

  11. Integral fast reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  12. Integral Fast Reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  13. The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1989-01-01

    In addition to maintaining the viability of its present commercial nuclear technology, a principal challenge in the US in the 1990s and beyond will be to regain and maintain a position among the world leadership in advanced reactor research and development. In this paper we'll discuss factors which we believe should today provide the rationale and focus for advanced reactor R and D, and we will then review the status of the major US effort, the Integral Fast Reactor (IFR) program

  14. The experimental nuclear reactor: AQUILON

    International Nuclear Information System (INIS)

    Girard, Y.; Koechlin, J.C.; Moreau, J.M.

    1958-01-01

    'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [fr

  15. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  16. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Neil Todreas; Pavel Hejzlar

    2008-01-01

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  17. Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    1998-04-01

    Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

  18. Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

    Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997. Refs,figs,tabs.

  19. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress for the past twelve months and will be continued this fiscal year, from April 1982 through March 1983, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1981. The 1982 year budget for R and D work and for construction of a prototype fast breeder reactor MONJU is approximately 20 and 27 billion yen respectively, excluding wages for the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaged in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and six operational cycles at 75 MWt were completed in December 1981. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor was approved by the authorities concerned, and the licensing of the first step was completed at the end of 1981. Preliminary design studies of a large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  20. Integral data for fast reactors

    International Nuclear Information System (INIS)

    Collins, P.J.; Poenitz, W.P.; McFarlane, H.F.

    1988-01-01

    Requirements at Argonne National Laboratory to establish the best estimates and uncertainties for LMR design parameters have lead to an extensive evaluation of the available critical experiment database. Emphasis has been put upon selection of a wide range of cores, including both benchmark, assemblies covering a range of spectra and compositions and power reactor mock-up assemblies with diverse measured parameters. The integral measurements have been revised, where necessary, using the most recent reference data and a covariance matrix constructed. A sensitivity database has been calculated, embracing all parameters, which enables quantification of the relevance of the integral data to parameters calculated with ENDF/B-V.2 cross sections

  1. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    Horak, C.; Purvis, E. III

    2000-01-01

    An examination has been made of how the currently available computing capabilities could be used to reduce Liquid Metal Fast Reactor design, manufacturing, and construction cost. While the examination focused on computer analyses some other promising means to reduce costs were also examined. (author)

  2. Thermophysical properties of fast reactor fuel

    International Nuclear Information System (INIS)

    Fink, J.K.

    1984-01-01

    This paper identifies the fuel properties for which more data are needed for fast-reactor safety analysis. In addition, a brief review is given of current research on the vapor pressure over liquid UO 2 and (U,PU)O/sub 2-x/, the solid-solid phase transition in actinide oxides, and the thermal conductivity of molten urania

  3. Charging machine for a fast production reactor

    International Nuclear Information System (INIS)

    Artem'ev, L.N.; Kurilkin, V.V.

    1971-01-01

    Charging machine for a fast production reactor is described. The machine contains charging mechanism, mechanism for positioning fresh fuel and spent fuel assemtlies, storage drums with sockets for control rod assemtlies and collet tongs for control rods. Recharging is conducted by means of ramp channel

  4. Fast breeder reactor at Kalkar. Pt. 2

    International Nuclear Information System (INIS)

    Degen, G.

    1979-02-01

    After a brief description of the previous development of the case the legal decisions are documented and commented on. The concept of the then FDP-Minister of Economy of North Rhine Westphalia (Riemer, Pu-combustion plant) is presented and the prospects and risk for the fast breeder reactor after the 3. partial construction license are discussed. (orig./HP) [de

  5. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    1993-03-01

    The management of radioactive waste is one of the key issues in today's discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow 'burning' of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs

  6. Fast reactors in Russia: State of the art and trends of development

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Ashurko, Yu.M.; Zverev, K.V.; Oshkanov, N.N.; Korol'kov, A.S.; Filin, A.I.

    2002-01-01

    This status report contains the following: facts on nuclear power in Russia from 2001-2002; plans for further development of nuclear power; state of the art on operation of fast reactors in 2002, namely BN-600, experimental reactors BOR-60 and BR-10; construction of NPP BN-800; participation in activities on BN-350 reactor decommissioning; description of trends of design studies in the field of fast reactors and accelerator driven systems

  7. 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)

  8. [International Thermonuclear Experimental Reactor support

    International Nuclear Information System (INIS)

    Dean, S.O.

    1990-01-01

    This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to ''support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: ''to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to ''assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

  9. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Vautrey, L.

    1982-01-01

    Only minor changes were introduced in the French nuclear programme by the new government in 1981. The operating conditions of Rapsodie were very satisfactory up to January 1982. After a leak in the double primary jacket (nitrogen circuit) the reactor was shut down for investigations. Phenix is continuing to operate smoothly. Construction of Super Phenix (Creys Malville power plant) is proceeding normally though with some delay. The studies for the future (after Creys Malville) are following their way both for the Project 1500 (Super Phenix 2) and for the specific plants of the fuel cycle. Research and development are largely directed toward Super Phenix 1 needs and the prospects of Super Phenix 2. International cooperation remains very intensive. The financial resources devoted to the development of fast reactors are globally stable. Including fuel cycle and safety (but excluding the Phenix operation) about 1300 millions of francs will be devoted to fast reactors by the C.E.A. in 1982. (author)

  10. Nuclear data for advanced fast reactors

    International Nuclear Information System (INIS)

    Rabotnov, N.S.

    2001-01-01

    Interest revives to fast reactors as the only proven technology obviously able of satisfying human energy needs for the next millennium by using full energy content of both natural uranium resources and of vast stocks of depleted uranium. This interest stimulates revision and improvement of fast reactor ND. Progress in reactor calculations accuracy due to better codes and much faster computers also increases relative importance of the input data uncertainties, especially in case of small reactivity margin and fuels of equilibrium compositions. The main objects of corresponding R and D efforts should be minor actinides and heavy liquid metal coolant. Data error bands and covariance information also gain importance as necessary components of neutron physics calculations. (author)

  11. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Xu Mi

    2000-01-01

    Considering the future clean energy supply in China, a rather consistent opinion is to develop nuclear power step by step with the contribution from a supplementary one up to an important one. The large scale utilization of nuclear energy obviously determines the interest in fast breeders; China right now already has about 300 GWe total electricity capacity using conventional energy resources. As the first step for fast reactor technology development in the country, the China Experimental Fast Reactor (CEFR) project is still under detail design stage, which is a sodium cooled pool type fast reactor with 65 MW thermal power matched with a turbine-generator of 25 MW. The ordering of the components is continuing. The site is ready and the steel works for the 3 m x 69 m x 82.5 m foundation base of reactor building are being arranged layer by layer. The review to the PSAR by the China National Nuclear Safety Administration (CNNSA) is going to the final stage, if everything goes smoothly. The first pouring of the concrete for the reactor building will be in the middle of the year 2000. The brief introduction of the CEFR design, safety characteristics, the main results of the safety analysis and design test demonstration are given in the paper. (author)

  12. Fast breeder reactor electromagnetic pump

    International Nuclear Information System (INIS)

    Araseki, Hideo; Murakami, Takahiro

    2008-01-01

    Main pumps circulating sodium in the FBR type reactor have been mechanical types, not electromagnetic pumps. Electromagnetic pump of 1-2 m 3 /min has been used as an auxiliary pump. Large sized electromagnetic pumps such as several hundred m 3 /min have not been commercialized due to technical difficulties with electromagnetic instability and pressure pulsations. This article explained electromagnetic and fluid equations and magnetic Reynolds number related with electromagnetic pumps and numerical analysis of instability characteristics and pressure pulsations and then described applications of the results to FBR system. Magnetic Reynolds number must be chosen less than one with appropriate operating frequency and optimum slip of 0.2-0.4. (T. Tanaka)

  13. Axial distribution of absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.; Ninkovic, M.

    1988-11-01

    The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

  14. Utilization of MVP for research on fast reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji

    2001-01-01

    Utilization of the continuous energy Monte-Carlo code, MVP, for research on fast reactor in Power Reactor and Nuclear Fuel Development Corporation(PNC) is described. In this report, three types of utilization are reviewed; (1) a comparison of the eigenvalues calculated by MVP with the results by the deterministic methods, (2) an improvement of U-238 reaction rate evaluation in JUPITER experimental Analysis and (3) an evaluation of heterogeneity effects for Am reaction rates of the moderated subassemblies. Since the results of MVP can be used as references, MVP is very useful code in research on fast reactor. It is one of indispensable tools in order to verify the models in the deterministic methods. Furthermore, it can be used so as to investigate the new concept reactors, such as a reactor aiming to transmute minor actinides(MA). On the other hand, a problem of the variance reduction remains. Especially, a small reactivity cannot be estimated by MVP because of large variances. The development of a Monte-Carlo method for a small reactivity calculation will promote the utilization of MVP for research on fast reactor. (author)

  15. Breeding description for fast reactors and symbiotic reactor systems

    International Nuclear Information System (INIS)

    Hanan, N.A.

    1979-01-01

    A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

  16. 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)

  17. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  19. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  20. Intrinsically secure fast reactors with dense cores

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2007-01-01

    Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: ·Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. ·Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total power. The modules can be used also independently (as

  1. Integral fast reactor shows its mettle

    International Nuclear Information System (INIS)

    Chang, Ya.; Lajnberri, M.; Barris, L.; Uoters, L.

    1988-01-01

    The main aspects of the problem of developing a closed fuel cycle at a NPP built in the so-called integrated version when a fast reactor and the plant for spent fuel regeneration and fuel element production are located in the same site (IFR project), are considered. The technologies of U-Pu-Zr alloy fuel reprocessing and production based on high-temperature metallurgical process and the method of casting under pressure are described. The demonstration of practical feasibility of the fuel cycle on the basis of the IFR reactor is planned for 1990

  2. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Johnson, R.N.

    1984-04-01

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  3. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  4. EDF research on fast neutron reactors

    International Nuclear Information System (INIS)

    In order to make possible the calculation of the temperatures of the sodium, of the sheath and of the fuel in fast reactor assemblies, taking into account the mixing phenomena induced by the helicoidal wires, two design codes have been developed. Those codes have then been adapted for their integration in the Superalcyon system. This system shall constitute the reference tool for the development of those codes that shall manage Phenix, and other reactors of the family. Cooling accidents, thermohydraulic studies, and steam generator studies are also in progress

  5. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Picker, C.; Ainsworth, K.F.

    1996-01-01

    The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

  6. A review of the UK fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Picker, C [AEA Technolgy plc, Risley, Warrington, Cheshire (United Kingdom); Ainsworth, K F [British Nuclear Fuels plc, Sellafield, Cumbria (United Kingdom)

    1996-07-01

    The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

  7. Material choices for the commercial fast reactor steam generators

    International Nuclear Information System (INIS)

    Willby, C.; Walters, J.

    1978-01-01

    Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

  8. The fast reactor and electricity supply, a utility view

    International Nuclear Information System (INIS)

    Wright, J.K.; Hall, R.S.; Kemmish, W.B.; Thorne, R.T.

    1982-01-01

    The significance of the fast reactor is discussed from the viewpoint of the Central Electricity Generating Board. The need for the fast reactor and a possible timescale for its introduction are examined. It is emphasised that demonstration of the commercial and environmental acceptability of the fuel cycle will be needed before any commitment can be made to fast reactors. (U.K.)

  9. What is the future for fast reactor technology?

    International Nuclear Information System (INIS)

    Kraev, Kamen

    2017-01-01

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  10. What is the future for fast reactor technology?

    Energy Technology Data Exchange (ETDEWEB)

    Kraev, Kamen [NucNet, Brussels (Belgium). The Independent Global Nuclear News Agency

    2017-08-15

    NucNet spoke to Vladimir Kriventsev, team leader for fast reactor technology development at the International Atomic Energy Agency (IAEA), about the possibilities and challenges of technology development in the fast reactor sector. Today, the field of fast reactors is vibrant and full of fascinating developments, some which will have an impact in the nearer term and others in the longer term.

  11. Review of fast reactor activities in India (1982-83)

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1983-01-01

    A review of fast reactor activities in India in 1982-1983 is given. One stage of construction of Fast Breeder Test Reactor (FBTR) is briefly described. The emphasis is on design studies for the 500 MWe Prototype Fast Breeder Reactor (PFBR). The main features of this design are introduced

  12. A review of fast reactor progress in Japan, March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K

    1979-07-01

    The fast reactor development project in Japan will be continued in the next fiscal year, from April 1979 through March 1980, at a similar scale of effort both in budget and personnel, to those of the fiscal year of 1978. The total budget for LMFBR development for the next fiscal year is approximately 24 billion Yen, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast reactor development in the PNC is approximately 500, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, approval for power increase from presently approved 50 MWt to 75 MWt with the present core and also to 100 MWt with a modified core in the future was granted by the regulatory authority in September 1978. Two operational cycles at 50 MWt have been completed very recently and preparation for power increase to 75 MWt is being made. With respect to the prototype fast breeder reactor MONJU, progress toward construction is being made and an environmental impact statement of MONJU filed last autumn is being reviewed by the concerned authorities. By the new atomic energy law recently made effective in Japan, the tasks of the former Japan Atomic Energy Commission were split into two and the Atomic Energy Safety Commission was newly established on 4th October 1978 in order to deal with nuclear safety problems in the country. All other problems are treated by the Atomic Energy Commission, as before. Highlights and topics of the fast reactor development activities in the past twelve months are summarized in this paper.

  13. Argentinean activities related to Fast Reactors

    International Nuclear Information System (INIS)

    Azpitarte, Osvaldo

    2012-01-01

    CNEA objectives in the area of Generation IV nuclear reactors: Implement a programme for the monitoring of the global progress of new technologies for Generation IV nuclear reactors and their fuel cycles, in order to generate and assess associated lines of R&D. – Perform studies and evaluations for defining the Generation IV line or lines on which CNEA would be interested; – Promote the participation on specific international projects; – Implementation of experimental facilities

  14. Chemical surveillance of commercial fast breeder reactors

    International Nuclear Information System (INIS)

    Stamm, H.H.; Stade, K.Ch.

    1988-01-01

    After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

  15. Fast reactor versions: elements of choice

    International Nuclear Information System (INIS)

    Tassart, J.; Zerbib, J.C.

    1984-01-01

    This paper has the objective of explaining in detail the economical, political, social and technical elements on which the CFDT (French Trade Union) bases its opposition to the commercial development of the version of fast reactors. An examination of the different choices which were investigated does not point to any legitimate grounds for this choice. What has to be done is to present the facts which enable the greatest possible number of workers or civilians to take up a position on the choices concerning them. A technical comparison of the fast neutron reactor with those operating at present is put forward (France and United Kingdom). It covers the different radioactive waste products and the results of the individual and collective monitoring of the workmen [fr

  16. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Shoai Tehrani, Bianka; Da Costa, Pascal

    2013-01-01

    Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

  17. Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

    International Nuclear Information System (INIS)

    Kamide, H.; Hayashi, K.; Momoi, K.

    1997-01-01

    The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

  18. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  19. Non-electric Applications of Fast Reactors

    International Nuclear Information System (INIS)

    Safa, Henri; Borgard, Jean-Marc

    2013-01-01

    Conclusions: → Most of industrial applications (80%) require low temperature heat below 540°C; → Fast Reactors are technically suitable to provide industrial steam at temperatures not accessible by standard LWRs; → As an illustrative example, the application at an oil refinery site has been studied showing the economic benefits; → Nuclear Cogeneration enhances the overall energy efficiency of the power plant; • Nuclear Cogeneration allows massive cut in CO 2 emissions

  20. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  1. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  2. Nodal method for fast reactor analysis

    International Nuclear Information System (INIS)

    Shober, R.A.

    1979-01-01

    In this paper, a nodal method applicable to fast reactor diffusion theory analysis has been developed. This method has been shown to be accurate and efficient in comparison to highly optimized finite difference techniques. The use of an analytic solution to the diffusion equation as a means of determining accurate coupling relationships between nodes has been shown to be highly accurate and efficient in specific two-group applications, as well as in the current multigroup method

  3. Some aspects of fast reactor economics

    International Nuclear Information System (INIS)

    Kazachkovskij, O.D.

    1996-01-01

    Expedient approach to evaluation of economic efficiency of fast reactors is discussed. It is concluded that determination of electric power generation cost should be based on the fact, that plutonium cost is dictated only by expenses for its extraction from the spent fuel. The cost of the first critical load is not included into capital investments, and investment charges should be sufficiently lower, than standard ones. 5 refs

  4. Thermal baffle for fast-breeder reactor

    International Nuclear Information System (INIS)

    Rylatt, J.A.

    1977-01-01

    A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel. 3 claims, 2 figures

  5. Fast reactor operation in the United States

    International Nuclear Information System (INIS)

    Smith, R.R.; Cissel, D.W.

    1978-01-01

    Of the many American facilities dedicated to fast reactor technology, six qualify as liquid-metal-cooled fast reactors. All of these satisfy the following criteria: an unmoderated neutron spectrum, highly enriched fuel material, substantial heat production, and the use of a liquid metal coolant. These include the following: EBR-I Clementine, LAMPRE, EBR-II, EFFBR, and SEFOR. Collectively, these facilities encompassed all of the more important features of liquid-metal-cooled fast reactor technology. Coolant types ranged from mercury in Clementine, to NaK in EBR-I, and sodium in the others. Fuels included enriched-uranium metallic alloys in EBR-I, EBR-II, and EFFBR; metallic plutonium in Clementine; molten plutonium alloy in LAMPRE; and a mixed UO 2 -PuO 2 ceramic in SEFOR. Heat removal techniques ranged from air-blast cooling in LAMPRE and SEFOR; steam-electrical generation in EBR-I, EBR-II, and EFFBR; to a mercury-to-water heat dump in Clementine. Operational experience with such diverse systems has contributed heavily to the U.S. Each of the six systems is described from the viewpoints of purpose, history, design, and operation. Attempts are made to limit descriptive material to the most important features and to refer the reader to a few select references if additional information is needed

  6. The integral fast reactor - an overview

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Hannum, W.H.

    1997-01-01

    The Integral Fast Reactor (IFR) is a system that consists of a fast-spectrum nuclear reactor that uses metallic fuel and liquid-metal (sodium) cooling, coupled with technology for high-temperature electrochemical recycling, and with processes for preparing wastes for disposition. The concept is based on decades of experience with fast reactors, adapted to priorities that have evolved markedly from those of the early days of nuclear power. It has four essential, distinguishing features: efficient use of natural resources, inherent safety characteristics, reduced burdens of nuclear waste, and unique proliferation resistance. These fundamental characteristics offer benefits in economics and environmental protection. The fuel cycle never involves separated plutonium, immediately simplifying the safeguarding task. Initiated in 1984 in response to proliferation concerns identified in the International Nuclear Fuel Cycle Evaluation (INFCE, 1980), the project has made substantial technical progress, with new potential applications coming to light as nuclear weapons stockpiles are reduced and concerns about waste disposal increase. A breakthrough technology, the IFR has the characteristics necessary for the next nuclear age. (author)

  7. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Xu, M.

    2002-01-01

    China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basic strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan (2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before 2010 respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor (CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached 16.8m above the ground. Forty seven components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which is only under consideration up to now. Some important technical selections have been settled, but its design has not yet started. (author)

  8. The Progress of Fast Reactor Technology Development in China

    International Nuclear Information System (INIS)

    Yang, Hongyi; Xu, Mi

    1994-01-01

    China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basis strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan(2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor(CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached the top namely 57m above the ground. More than one hundred components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m 2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which in only under consideration up to now. Some important technical selections have been settled, but its design has not yet started

  9. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Wild, E.; Mack, K.J.; Gegenheimer, M.

    1984-11-01

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.) [de

  10. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  11. Today's attitudes and future prospects of fast reactors in Italy

    International Nuclear Information System (INIS)

    Barabaschi, S.; Cicognani, G.; Pierantoni, F.

    1982-01-01

    The Italian fast reactor programme is reviewed. The 15 year collaboration with France has resulted in the construction of the PEC reactor, development of the Superphenix-1 and a common R and D programme for future large fast reactors. The CNEN 4th five year (1980-84) plan is outlined. The budget breakdown for different areas shows the importance attached to the fast reactor. (U.K.)

  12. Substantiation of physical concepts of fast reactors in Russia: experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

    2008-07-01

    The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

  13. Kinetic studies on a repetitively pulsed fast reactor

    International Nuclear Information System (INIS)

    Das, S.

    1982-01-01

    Neutronic analysis of an earlier proposed periodically pulsed fast reactor at Kalpakkam (KPFR) has been carried out numerically under equilibrium and transient conditions using the one-point model of reactor kinetics and the experimentally measured total worth of reactivity modulator, the parabolic coefficient of reactivity of the movable reflector and the mean prompt neutron lifetime. Results of steady-state calculations - treated on the basis of delayed neutron precursor and energy balances during a period of operation - have been compared with the analytical formulae of Larrimore for a parabolic reactivity input. Empirical relations for half-width of the fast neutron pulse, the peak pulse power and the power at first crossing of prompt criticality have been obtained and shown to be accurate enough for predicting steady-state power pulse characteristics of a periodically pulsed fast reactor. The concept of a subprompt-critical reactor has been used to calculate the fictitious delayed neutron fraction, β of the KPFR through a numerical experiment. Relative pulse height stability and pulse shape sensitivity to changes of maximum reactivity is discussed. With the aid of new safety concepts, the Power Amplification Factor (PAF) and the Pulse Growth Factor (Rsub(p)), the dynamics KPFR under accidental conditions has been studied for step and ramp reactivity perturbations. All the analysis has been done without taking account of reactivity feedback. (orig.)

  14. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    1998-01-01

    This report describes the development and activities on fast reactor in Japan for the period of April 1996 - March 1997. During this period, the 30th duty cycle operation has been started in the Experimental Fast Reactor ''''Joyo''''. The cause investigation on the sodium leak incident has completed and the safety examination are being performed in the Prototype Fast Breeder Reactor ''''Monju''''. The three years design study since FY1994 on the plant optimization of the Demonstration FBR has been completed by the Japan Atomic Power Company (JAPC). Related research and development works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of Power Reactor and Nuclear Fuel Development Corporation (PNC), JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). In November 1996, the Japan Atomic Energy Commission (JAEC) established a Social Gathering Meeting to discuss generally the significance of FBR development in Japan for the future. (author)

  15. Materials development for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T.; Mathew, M.D.; Laha, K.; Sandhya, R., E-mail: san@igcar.gov.in

    2013-12-15

    Highlights: • A modified version of alloy D9 designated as IFAC-1 has been developed. • Oxide dispersion strengthened Grade 91 steel with good creep strength developed. • 0.14 wt% nitrogen in 316LN stainless steel leads to improved mechanical properties. • Type IV cracking resistant Grade 91 steel with boron addition developed. • Mechanical properties of SFR materials evaluated in sodium environment. -- Abstract: Materials play a crucial role in the economic competitiveness of electricity produced from fast reactors. It is necessary to increase the fuel burn-up and design life in order to realize this objective. The burnup is largely limited by the void swelling and creep resistance of the fuel cladding and wrapping materials. India's 500 MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction. The major structural materials chosen for PFBR with MOX fuel are D9 austenitic stainless steel as fuel clad and wrapper material, 316LN austenitic stainless steel for reactor components and piping and modified 9Cr-1Mo steel for steam generator. In order to improve the burnup, titanium, phosphorous and silicon contents in alloy D9 have been optimized for decreased void swelling and increased creep strength and this has led to the development of a modified version of alloy D9 as IFAC-1. Ferritic steels are inherently resistant to void swelling. The disadvantage is their poor creep strength. Creep resistance of 9Cr-ferritic steel has been improved with the dispersion of nano-size yttria to develop oxide dispersion strengthened (ODS) steel clad tube with long-term creep strength, comparable to alloy D9 so as to achieve higher fuel burnup. Improved versions of 316LN stainless steel with nitrogen content of about 0.14 wt% having higher creep strength to increase the life of fast reactors and modified 9Cr-1Mo steel with reduced nitrogen content and controlled addition of boron to improve type IV cracking resistance for steam generator

  16. Fast reactors: R and D targets and outlook for their introduction

    International Nuclear Information System (INIS)

    Poplavsky, V.; Barre, B.; Aizawa, K.

    1997-01-01

    In this paper the current status of fast reactors development is briefly outlined, including experimental, demonstration, and commercial installations. Data on the experience gained in development and operation of NPPs with reactors of this type are presented. The issues are discussed in connection with possibilities of fast reactor development in the nuclear power structure for the near (up to 2010-2020) and distant future. In the final part of the paper, an analysis is given of possible ways for R and D development in the field of NPPs with fast neutron reactors. (author)

  17. Methodology for substantiation of the fast reactor fuel element serviceability

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Maershin, A.A.

    1988-01-01

    Methodological aspects of fast reactor fuel element serviceability substantiation are presented. The choice of the experimental program and strategies of its realization to solve the problem set in short time, taking into account available experimental means, are substantiated. Factors determining fuel element serviceability depending on parameters and operational conditions are considered. The methodological approach recommending separate studing of the factors, which points to the possibility of data acquisition, required for the development of calculational models and substantiation of fuel element serviceability in pilot and experimental reactors, is described. It is shown that the special-purpose data are more useful for the substantiation of fuel element serviceability and analytical method development than unsubstantial and expensive complex tests of fuel elements and fuel assemblies, which should be conducted only at final stages for the improvement of the structure on the whole

  18. A review of fast reactor program in Japan - April 1984

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1984-01-01

    The fast breeder reactor development project in PNC has been in progress steadily in these eighteen years. Concerning the experimental fast reactor, JOYO, the MK-II core attained criticality on November 22, 1982 with 51 fuel assemblies, and received the ''Certificate of Inspection before Operation'' from Government Authority on March 31, 1983, after 100 hours operation with the rated output of 100 MW. Since then, the core has been utilized to implement irradiation bed characteristics test, and to irradiate fuels and structural materials especially for the prototype reactor MONJU. With respect to the prototype reactor MONJU, the installation permit was issued on May 27, 1983, from the prime minister, and the contracts of the first stage between PNC and fabricators were made recently. At the same time, almost all the licenses of preparatory construction works were issued by March 1983, and preparatory construction works were started in April 1983. On the other hand, conceptual design of a demonstration reactor is now under way in a close cooperation with concerned authorities and utilities, as well as investigations of the way of conducting necessary research and development

  19. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1980-01-01

    The performance test on the reactor power increase to 75 MW was started on July 3, and the target of 75 MW was reached on July 16, in the experimental fast reactor Joyo. The tests on the heat transport characteristics, power coefficient, the response to the change of outlet temperature, the loss of external power supply and so on were carried out, and the performance test was finished on August 23, except the test of 75 MW continuous operation. The annual inspection of the systems is being carried out in parallel with the regular inspection. The design to prepare for the manufacture of the prototype fast reactor Monju is being prepared. The analysis of decay heat removal is being carried out, and various calculation codes were developed. The technological survey on overseas LMFBRs is being made. The conceptual design of the demonstration reactor is being prepared. The research and development of reactor physics, structural components for Joyo and Monju, instrumentation and control, sodium technology, fuel materials, structural materials, safety problems and steam generators are reported. The tests on the transient boiling of sodium, fuel failure propagation, heat transfer between molten materials, post-accident decay heat removal and so on have been carried out. (Kako, I.)

  20. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

    2009-06-15

    This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

  1. The plutonium utilization in thermal and fast reactor in Japan

    International Nuclear Information System (INIS)

    Amanuma, T.; Uematsu, K.

    1977-01-01

    The nuclear power development in Japan is rather extensive one, and the installed nuclear power capacity is expected to reach 49,000 MWe by 1985 and possibly to reach 170,000 MWe by 2000 according to a prediction. Currently istalled nuclear power is mainly based on Light Water type Reactor, and this trend is expected to persist for the time-being. The plutonium produced by LWR will be accumulated to 20 tons by 1985 and to more than 200 tons by 2000. If the produced plutonium will simply be stored, it will raise the economic pressure to utilities and the management and physical protection problems associated with plutonium storing. Therefore, it is not too wise simply to store plutonium in a locked vault. In Japan, there are three ways of solving these problems which are currently worked out. There is no doubt that the best solution is to use plutonium in fast reactors. To reach this goal, an Experimental Fast Reactor ''JOYO''has been constructed and it is waiting for criticality in very near future. A prototype fast breeder reactor ''MONJU'', which is designed for about 300 MWe, is nearing to the last stage of the design work. The start of its construction will take place in a few yesars. The domonstration fast breeder reactor will come next to ''MONJU'' and the large scale commercial use of fast breeder reactor is expected to start around 1995. To anwer the near-term need for plutonium utilization, two technologies, which are equally important to Japan, are currently developed. One is the recycle use of plutonium into LWR. This technology has long been jointly developed by research organizations and utilities. Some of fuel irradiation data are already obtained and the physics study has also been extensive. The application of this technology is expected to start about 1987. The other is to burn plutonium in an Advanced Thermal Reactor (D 2 O moderated, Boiling Water Cooled) which shows better characteristics of using plutonium. The 160 MWe ''Fugen'' is a prototype

  2. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    The fast reactor, which can generate electricity and breed additional fissile material for future fuel stocks, is a resource that will be needed when economic uranium supplies for the thermal reactors diminish. Further, the fast-fission fuel cycle in which material is recycled (a basic requirement to meet sustainability criteria) offers the flexibility needed to contribute decisively towards solving the problem of growing “spent” fuel inventories by greatly reducing the volume, the heat load and the radiotoxic inventory of high-level wastes that must be disposed of in long-term geological repositories. This is a waste management option that will play an increasingly important role in the future, and help to ensure that nuclear energy remains a sustainable long-term option in the world’s overall energy mix. In recognition of the fast reactor’s importance for the sustainability of the nuclear option, currently there is worldwide renewed interest in fast reactor technology development, as indicated, e.g., by the outcome of the Generation IV International Forum (GIF) technology review, which concluded with 3 out of 6 innovative systems to be fast reactors (gas cooled fast reactor, sodium cooled fast reactor, and heavy liquid metal cooled fast reactor), plus a potential fast core for a 4th concept, the super-critical water reactor. Currently, fast reactor construction projects are ongoing in India (PFBR) and Russian Federation (BN-800), whilst in China the first experimental fast reactor (CEFR) is in the commissioning phase. Fast reactor programs are also carried out in Europe (in particular in France), Japan, Republic of Korea and the USA. The most important challenges for fast reactors are in the areas of cost competitiveness with respect to LWRs and other energy sources, enhanced safety, non-proliferation, and public acceptance. With the exception of this latter, these translate into technology development challenges, i.e. the development of advanced reactor

  3. Gas cooled fast breeder reactors using mixed carbide fuel

    International Nuclear Information System (INIS)

    Kypreos, S.

    1976-09-01

    The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

  4. Fast Reactor Programme. Third Quarter 1969. Progress Report

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1970-02-01

    The RCN research programme on fast spectrum nuclear reactors comprises reactor physics, fuel performance, radiation damage in canning materials, corrosion behaviour in canning materials, aerosol research and heat transfer and hydraulics. An overview is given of the fast reactor experiments at the STEK critical facility in Petten, the Netherlands, in the third quarter of 1969

  5. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

    2010-02-01

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  6. Superalloy applications in the fast breeder reactor

    International Nuclear Information System (INIS)

    Powell, R.W.

    1976-01-01

    The economics of the LMFBR are dependent on the breeding of new fuel in the reactor core and this can be improved by the use of advanced alloys as core structural components. The environment of the core makes superalloys a natural choice for these components, but phenomena related directly to neutron irradiation necessitate extensive testing. Consequently, commercially-available superalloys, together with a number of developmental alloys are being tested in existing LMFBR's and by simulation techniques to determine the best alloy for use in the LMFBR core. It presently appears that such materials will indeed be capable of the performance required, and will greatly facilitate the commercial realization of the fast breeder reactor

  7. The economics of fast breeder reactors

    International Nuclear Information System (INIS)

    Rapin, M.

    1990-01-01

    The overall status of the fast breeder reactor (FBR) system is periodically reviewed in France. In 1983, a report was prepared on the status and prospects of the FBR system at the request of the then Minister of Industry. Five years later, Electricite de France (EdF) and the French Atomic Energy Commission (CEA) jointly updated this report. The FBR reactor system economic considerations mentioned here are taken from the work performed in 1987-88 for this updating. The position in 1983 is reviewed to highlight concrete developments. Developments that have occurred since then are presented, along with the prospects that today enable us to define better the technical and economic potential of the FBR system. In conclusion, the effects of these findings on desirable directions are discussed, in particular with regard to European FBR cooperation. (author)

  8. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

  9. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

  10. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    Dreimanis, Andrejs

    2015-01-01

    An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

  11. GENIUS & the Swedish Fast Reactor programme

    International Nuclear Information System (INIS)

    Wallenius, Janne

    2012-01-01

    Concluding remarks: Sweden’s growing fast reactor programme focuses on LFR technology, but we also participate in ASTRID. • An innovative facility for UN fabrication, an LBE thermal hydraulics loop and a lead corrosion facility are operational. • A plutonium fuel fabrication lab is is under installation (this week!) • The government is assessing the construction of ELECTRA-FCC, a centre for Gen IV-system R&D, at a tentative cost of ~ 140±20 M€. • Location: Oskarshamn (adjacent to intermediate repository) • Date of criticality: 2023 (best case) • Swedish participation in IAEA TWG-FR should intensify

  12. Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Levinskas, D.

    1993-01-01

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves

  13. Heterogeneous cores for fast breeder reactor

    International Nuclear Information System (INIS)

    Schroeder, R.; Spenke, H.

    1980-01-01

    Firstly, the motivation for heterogeneous cores is discussed. This is followed by an outline of two reactor designs, both of which are variants of the combined ring and island core. These designs are presented by means of figures and detailed tables. Subsequently, a description of two international projects at fast critical zero energy facilities is given. Both of them support the nuclear design of heterogeneous cores. In addition to a survey of these projects, a typical experiment is discussed: the measurement of rate distributions. (orig.) [de

  14. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

    1974-01-01

    The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

  15. Nuclear fuel assembly for fast neutron reactors

    International Nuclear Information System (INIS)

    Ilyunin, V.G.; Murogov, V.M.; Troyanov, M.F.; Rinejskij, A.A.; Ustinov, G.G.; Shmelev, A.N.

    1982-01-01

    The fuel assembly of a fast reactor consists of fuel elements comprising sections with fissionable and breeding material and tubes with hollows designed for entrapping gaseous fission products. Tubes joining up to the said sections are divided in a middle and a peripheral group such that at least one of the tube groups is placed in the space behind the coolant inlet ports. The configuration above allows reducing internal overpressure in the fuel assembly, thus reducing the volume of necessary structural elements in the core. (J.B.)

  16. The Argentine-Brazilian fast reactor programme

    International Nuclear Information System (INIS)

    Gho, C.J.; Mauricio, A.

    1989-01-01

    This paper summarizes the Argentine-Brazilian Fast Reactor Programme and gives reasons for the decision of a binational venture. The work carried out by both countries is described, showing how they complement each other, with the corresponding saving of resources. The main objectives of the Programme and tentative schedules in three progressing integrating stages are given and the present nuclear know-how in each country is identified as a good starting point. The paper also gives some details regarding the economical and human resources involved. (author). 1 graph

  17. Accident analysis for US fast burst reactors

    International Nuclear Information System (INIS)

    Paternoster, R.; Flanders, M.; Kazi, H.

    1994-01-01

    In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards

  18. Reprocessing of fast neutron reactor fuel

    International Nuclear Information System (INIS)

    Bourgeois, M.

    1981-05-01

    A PUREX process specially adapted to fast neutron reactor fuels is employed. The results obtained indicate that the aqueous process can be applied to this type of fuel: almost 10 years operation at the AT 1 plant which processes fuel from RAPSODIE; the good results obtained at the MARCOULE pilot plant on large batches of reference fuels. The CEA is continuing its work to transfer this technology onto an industrial scale. Industrial prototypes and the launching of the TOR (traitement d'oxydes rapides) project will facilitate this transfer. In 1984, it is expected that fast fuels will be able to be processed on a significant scale and that supplementary R and D facilities will be available [fr

  19. Benchmark test of evaluated nuclear data files for fast reactor neutronics application

    International Nuclear Information System (INIS)

    Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki

    2007-07-01

    A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)

  20. Review of fast reactor activities in Italy, April 1978

    Energy Technology Data Exchange (ETDEWEB)

    Pierantoni, F [CNEN Fast Reactor Programme, Bologna (Italy)

    1978-07-01

    In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments.

  1. Review of fast reactor activities in Italy, April 1978

    International Nuclear Information System (INIS)

    Pierantoni, F.

    1978-01-01

    In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments

  2. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  3. Development, operational experience and implications for future design of fast reactors in Western Europe

    International Nuclear Information System (INIS)

    Brandstetter, A.; Broomfield, A.M.; Saitcevsky, B.

    1990-01-01

    Over the past 30 years, the partners now collaborating in Europe on fast-reactor development have taken the technology from a theoretical possibility to an engineering reality. In that time there has been a progression from experimental zero-energy facilities followed by small power-producing engineering test reactors, to prototype reactors and a large commercial-size reactor. The paper describes the highlights of the reactor programmes in the partner countries and by example illustrates the experience gained from reactor operation and some of the principal activities in the supporting development programme. These include such topics as fuel performance, fast-neutron physics, liquid-metal thermal hydraulics, sodium chemistry, instrumentation and safety aspects. The paper concludes by summarizing some of the main objectives of the current development programme, which is directed to the support of the European Fast Reactor design being prepared by the European design and construction companies. (author)

  4. A small modular fast reactor as starting point for industrial deployment of fast reactors

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Lo Pinto, Pierre; Konomura, Mamoru

    2006-01-01

    The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO 2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

  5. 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)

  6. Status of fast breeder reactor development in Germany

    International Nuclear Information System (INIS)

    Hueber, R.; Kathol, W.; Kempken, M.

    1991-01-01

    The KNK, the sodium cooled compact reactor is an experimental nuclear power plant of 20 MW electric power. Since 1977, it has been operated with fast reactor cores as KNK II. The KNK II/3 core was designed. The core fabrication has been largely completed. In 1990, the KNK II plant achieved a time availability of 56%. On January 8, 1991 KNK II was shut down for inspection. Since pre-nuclear commissioning was completed the Kalkar Nuclear Power Station SNR 300 has been operated in a mode similar to that of a power station. In March 1991 the financing partners decided not to prolong the standby phase because they do not think that the last construction permit and the operation permit will be issued within a definite period of time. The partners were convinced that the lack of progress in the licensing procedure was not caused by basic safety deficiencies of the project but by the way the licensing procedure was executed. The German fast breeder programme is now concentrated on contributions to the European Fast Reactor. (author)

  7. 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.

  8. Status of National Programmes on Fast Breeder Reactors. International Working Group on Fast Reactors Twenty-First Annual Meeting, Seattle, USA, 9-12 May 1988

    International Nuclear Information System (INIS)

    1988-11-01

    The following papers on the status of national programmes on fast breeder reactors are presented in this report: Fast breeder reactor development in France during 1987; Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands; A review of the Indian fast reactor programme; A review of the Italian fast reactor programme; A review of the fast reactor programme in Japan; Status of fast reactor activities in the USSR; A review of the United Kingdom fast reactor programme; Status of liquid metal reactor development in the United States of America; Review of activities of the Commission of European Communities relating to fast reactors in 1987; European co-operation in the field of fast reactor research and development — 1987 progress report; A review of fast reactor activities in Switzerland

  9. Seminar on Heat-transfer fluids for fast neutron reactors

    International Nuclear Information System (INIS)

    Brechet, Yves; Dautray, Robert; Friedel, Jacques; Brezin, Edouard; Martin, Georges; Pineau, Andre; Carre, Francois; Gauche, Francois; Rodriguez, Guillaume; Latge, Christian; Cabet, Celine; Garnier, Jean-Claude; Bamberger, Yves; Sauvage, Jean-Francois; Buisine, Denis; Agostini, Pietro; Ulyanov, Vladimir; Auger, Thierry; Heuer, Daniel; Ghetta, Veronique; Bubelis, Evaldas; Charlaix, Elisabeth; Barrat, Jean-Louis; Boquet, Lyderic; Glickman, Evgueny; Escaravage, Claude

    2014-03-01

    This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

  10. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

  11. A review of fast reactor program in Japan - April 1983

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1983-01-01

    The fast breeder reactor development project in Japan has been in progress during the past twelve months and will be continued in the next fiscal year, from April 1983 through March 1984, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1982. Concerning the experimental fast reactor, JOYO, all the scheduled testings and operations were completed by the end of 1981 and from the beginning of 1982 the change-out work from Mark-I core to Mark-II core has been continued for 11 months. The initial criticality on the Mark-II core was achieved on 22 Nov. 1982 and after 3 months low power physics tests the reactor power was raised up to 100% (100 MWt) in the middle of March 1983. With respect to the prototype reactor MONJU, progress toward construction has been made and the licensing of the second step will be completed in the first half of 1983. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  12. Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

    1991-01-01

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

  13. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  14. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

    International Nuclear Information System (INIS)

    Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

    2010-01-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

  15. Interim waste storage for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

    1991-01-01

    The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

  16. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1979-01-01

    The experimental fast reactor ''Joyo'' will be tested at 75 MW output, starting in April, 1980. In connection with the accident in the Three Mile Island plant, the reexamination of the plant safety and the rechecking-up of the maintenance control system were carried out, and the special inspection by the Science and Technology Agency was executed from May 21 to 23, 1979. Thereafter, the preparation for raising the power output was completed. The periodical inspection after the completion of 50 MW operation is being carried out. The state of progress of various equipments and the codes for core characteristic analysis is reported. The construction preliminary design (2) of the prototype reactor ''Monju'' is examined, and the same design (3) is prepared. The analysis of the decay heat in the prototype reactor is carried on for the safety licensing. The technological investigation of LMFBRs in foreign countries is under way. The preliminary design (4) of the demonstration reactor is under examination, and the technical specifications of the conceptual design (1) are prepared. The researches and developments of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and materials, safety and steam generators are reported. (Kako, I.)

  17. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    International Nuclear Information System (INIS)

    Moiseyev, A.V.

    2008-01-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  18. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  19. Fast reactor knowledge preservation system: Taxonomy and basic requirements

    International Nuclear Information System (INIS)

    2008-01-01

    The IAEA has taken the initiative to coordinate efforts of Member States in the preservation of knowledge in the area of fast reactors. In the framework of this initiative, the IAEA intends to create an international database compiling information from different Member States on fast reactors through a web portal. Other activities related to this initiative are being designed to accumulate and exchange information on the fast reactor area, to facilitate access to this information by users in different countries and to assist Member States in preserving the experience gained in their countries. The purpose of this publication is to develop a taxonomy of the Fast Reactor Knowledge Preservation System (FRKPS) that will facilitate the preservation of the world's fast reactor knowledge base, to identify basic requirements of this taxonomy on the basis of the experience gained in the fast reactor area, as well as results of previous IAEA activities on fast reactor knowledge preservation. The need for such taxonomy arises from the fact that during the past 15 years there has been stagnation in the development of fast reactors in the industrialized countries that were involved, earlier, in intensive development of this area. All studies on fast reactors have been stopped in countries such as Germany, Italy, the United Kingdom and the United States of America and the only work being carried out is related to the decommissioning of fast reactors. Many specialists who were involved in the studies and development work in this area in these countries have already retired or are close to retirement. In countries such as France, Japan and the Russian Federation that are still actively pursuing the evolution of fast reactor technology, the situation is aggravated by the lack of young scientists and engineers moving into this branch of nuclear power

  20. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  1. Actinide behavior in the Integral Fast Reactor. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  2. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  3. Behavior of actinides in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Courtney, J.C.; Lineberry, M.J.

    1994-01-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  4. Steam generator tubing development for commercial fast breeder reactors

    International Nuclear Information System (INIS)

    Sessions, C.E.; Uber, C.F.

    1981-01-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs

  5. Study of the seismic behaviour of the fast reactor cores

    International Nuclear Information System (INIS)

    Cerqueira, E.

    1998-01-01

    This work studies the seismic behaviour of fast neutrons reactor cores. It consists in analyzing the tests made on the models Rapsodie and Symphony by using the calculation code Castem 2000. Te difficulty is in the description of connections of the system and the effects of the fluid (calculation in water). The results for the programme Rapsodie are near the experimental results. For the programme Symphony, the calculations in air have allowed to represent the behaviour of fuel assemblies in a satisfying way. It is still to analyze the tests Symphony in water. (N.C.)

  6. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

    The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

  7. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  8. Improvement of covariance data for fast reactors

    International Nuclear Information System (INIS)

    Shibata, Keiichi; Hasegawa, Akira

    2000-02-01

    We estimated covariances of the JENDL-3.2 data on the nuclides and reactions needed to analyze fast-reactor cores for the past three years, and produced covariance files. The present work was undertaken to re-examine the covariance files and to make some improvements. The covariances improved are the ones for the inelastic scattering cross section of 16 O, the total cross section of 23 Na, the fission cross section of 235 U, the capture cross section of 238 U, and the resolved resonance parameters for 238 U. Moreover, the covariances of 233 U data were newly estimated by the present work. The covariances obtained were compiled in the ENDF-6 format. (author)

  9. Trial visualization of fast reactor design knowledge

    International Nuclear Information System (INIS)

    Yoshikawa, Shinji; Minami, Masaki; Takahashi, Tadao

    2011-01-01

    In design problems of large-scale systems like fast breeder reactors, inter-relations among design specifications are very important where a selected specification option is transferred to other specification selections as a premise to be taken account in engineering judgments. These inter-relations are also important in design case studies with the hypothetical adoption of rejected design options for the evaluation of deviation propagations among design specifications. Some of these rejected options have potential worth for future reconsideration by some circumstance changes (e.g., advanced simulations to exclude needs for mock-up tests, etc.), to contribute to flexibility in system designs. In this study, a computer software is built to visualize a design problem structure by representing engineering knowledge nodes on individual specification selections along with inter-relations of design specifications, to validate the knowledge representation method and to derive open problems. (author)

  10. Progress report on fast breeder reactor development at PNC, Japan, October - December, 1974

    International Nuclear Information System (INIS)

    1975-03-01

    Following the completion of building construction and equipment installation for the experimental fast breeder reactor ''Joyo'' at PNC's Oarai Engineering Center, hydrostatic pressure and leak tests were conducted on the reactor vessel. For the prototype fast breeder reactor ''Monju'', specification was finalized after the design adjustment. For the period from October to December, 1974, the following matters are described: construction of the Joyo, design of the Monju, reactor physics, components and equipments, instruments and control, sodium technology, fuel and material research and development, safety research and development, and steam generator. (Mori, K.)

  11. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

    Reference is made to liquid metal cooled fast breeder reactors of the 'pool' kind. In this type of reactor the irradiated fuel is lowered into a transfer rotor for removal to storage facilities, this rotor normally having provision for the temporary storage of 20 irradiated fuel assemblies, each within a stainless steel bucket. For insertion or withdrawal of a fuel assembly the rotor is rotated to bring the fuel assembly to a loading or discharging station. The irradiated fuel assembly is withdrawn from the rotor within its bucket and the total weight is approximately 1000 kg, which is lifted about 27 m. In the event of malfunction the combination falls back into the rotor with considerable force. In order to prevent damage to the rotor fracture pins are provided, and to prevent damage to the reactor vessel and other parts of the reactor structure deformable energy absorbing devices are provided. After a malfunction the fractured pins and the energy absorbing devices must be replaced by remote control means operated from outside the reactor vault - a complex operation. The object of the arrangement described is to provide improved energy absorbing means for fuel assemblies falling into a fuel transfer rotor. The fuel assemblies are supported in the rotor by elastic means during transfer to storage and a hydraulic dash pot is provided in at least one position below the rotor for absorbing the energy of a falling fuel assembly. It is preferable to provide dash pots immediately below a receiving station for irradiated fuel assemblies and immediately below a discharge station. Each bucket is carried in a container that is elastically supported in the transfer rotor on a helical coil compression spring, so that, in the event of a malfunction the container and bucket are returned to their normal operating position after the force of the falling load has been absorbed by the dash pot. The transfer rotor may also be provided with recoil springs to absorb the recoil energy

  12. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  13. Measurement of fast neutron spectra inside reactors with a Li{sup 6} semiconductor counter spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, V S; Lalovic, B I; Petrovic, B P [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1963-12-15

    The possibility of using the Li{sup 6} semiconductor counter spectrometer for measuring fast neutron spectra inside reactors has been investigated in details and some solutions of the difficulties associated with the high interference of thermal neutrons in well-moderated reactors are suggested and checked experimentally (author)

  14. The experimental and technological developments reactor

    International Nuclear Information System (INIS)

    Carbonnier, J.L.

    2003-01-01

    THis presentation concerns the REDT, gas coolant reactor for experimental and technological developments. The specifications and the research programs concerning this reactor are detailed;: materials, safety aspects, core physic, the corresponding fuel cycle, the reactor cycle and the program management. (A.L.B.)

  15. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

    2011-01-01

    Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  16. Actinide burning in the integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1993-01-01

    During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as 99 Tc, 129 I, and 14 C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process

  17. Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

    International Nuclear Information System (INIS)

    Jammes, Ch.

    2010-07-01

    After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

  18. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  19. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  20. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H. (Argonne National Lab., IL (United States)); Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France)); Bagley, K.Q. (AEA Reactor Services, Risley (United Kingdom)); Crittenden, G.C. (AEA Reactor Services, Dounreay (United Kingdom)); Dievoet, J. van (Belgonucleaire, Brussels (Belgium))

    1993-09-01

    The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

  1. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Xu Mi

    1998-01-01

    The paper will outline the main activities on fast reactor technology in China. In the year 1996, with the increasing of about 15 GWe installed electricity capacity, the total national electricity generation capacity has reached 225 GWe in the Country. Two nuclear power plants, Qinshan Phase 1 and Daya Bay have their rather good operation. The load factor of Qinshan Phase 1 was 84.7%. 76.1% and 64.1% for Daya Bay Unit 1 and Unit 2 respectively. During the Ninth 5-year (from 1996 to 2000) four NPPs Consisting of eight units of installed 6620MWe will be constructed. Under the framework of the High Technology Programme the Chinese Experimental Fast Reactor (CEFR) with the power 65MWth matched with 25MWe turbine-generator is still under preliminary design stage, which is sodium cooled pool type, (Pu,U)O 2 as fuel, in-core primary Went fuel storage, two mechanical pumps and four intermediate heat exchangers for primary circuit two loops for secondary circuits two independent immersed heat exchangers and air coolers with high stacks for passive residual heat removal system. Some design changes are presented in the paper. Concerning the R and D for the CEFR, besides the facilities already prepared, for demonstration of thermohydraulic characteristics of natural convection, a water simulation reactor pool facility in about one third scale is planned, in order to prepare the reactor physics experiments for its start-up, the zero power fast neutron facility with 50kg U-235 has been restored, for endurance testing of core subassemblies and getting some sodium loop operation experiences, Italian ESPRESSO and CEDI are under reconstruction in our lab. As for the engineering preparation of the project CEFR, the Feasibility Study Report was approved by Authorities on November last year. The site preparation and the design of incorporated to grid are just started. Finally, the activities of the international cooperation are presented in the paper. (author)

  2. Fast reactor physics at CEA: present studies and future prospects

    International Nuclear Information System (INIS)

    Hammer, P.

    1980-09-01

    This paper aims at giving a general survey of the fast reactor core physics and shielding studies wich are in progress at CEA (1979-1983) in order to solve the neutronic problems related to: - core design optimization, - reactor operation and fuel management, - safety, for the development of fast commercial breeders in France after the SUPER-PHENIX 1 construction is achieved

  3. The problems of thermohydraulics of prospective fast reactor concepts

    International Nuclear Information System (INIS)

    Sedov, A.A.

    2000-01-01

    In this report the main requirements to fast reactors in system of future multicomponent Nuclear Power with closed U-Pu fuel cycle are regarded. The peculiarities of different liquid-metal (sodium and lead-alloyed) coolants as well as the thermohydraulics problems of prospective fast reactors (FR) concepts are discussed. (author)

  4. Programme and current status of fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    Suita, T.; Oyama, A.

    1977-01-01

    In 1967 the Japan Atomic Energy Commission revised her long term programme after a two year study for giving principles to her nuclear energy development programme, which indicated the dominant role of nuclear energy mid 1980's in the electric power generation and stressed the necessity of developing fast breeder reactors. It also recommended to organize a nucleus to undertake this nation-wide project, bringing together the total capability available throughout the country. Accordingly, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established in 1967 to develop two sodium-cooled fast reactors, an experimental fast reactor of about 100 MW thermal and a prototype fast breeder reactor of about 300 MW electrical, both using mixed oxide fuels. Construction of the experimental fast reactor started in 1970 and was essentially completed at the end of in 1974. The precommissioning test was followed in parallel with re-evaluating quality assurance of all systems. Physics test will be initiated around the end of 1976. The conceptual design of the prototype fast breeder reactor is now toward its final stage. Surveys on its proposed site have just started. Construction will start in 1978. Beside R and D works conducted by many organizations in Japan as well as under the international cooperation, several key test facilities were installed by PNC itself to conduct in-sodium test of full-size prototype components including 50 MW steam generators and post-irradiation-examination of fuels and materials. Recently an interim report was issued to an ad-hoc committee organized by JAEC to evaluate future prospect of the fuel cycle and power reactors. This recommended start of construction of the prototype reactor as scheduled and the large demonstration reactor to be followed to the prototype. Thus the fast breeder reactor is indicated as the most indispensable in 1990's

  5. MYRRHA – A multi-purpose fast spectrum research reactor

    International Nuclear Information System (INIS)

    Aït Abderrahim, Hamid; Baeten, Peter; De Bruyn, Didier; Fernandez, Rafael

    2012-01-01

    Highlights: ► Historical evolution of the MYRRHA project. ► Detail design of the MYRRHA Accelerator Driven System. ► Irradiation performance simulation of the MYRRHA ADS. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental Accelerator-Driven System (ADS) currently under development at SCK⋅CEN and will replace the Material Testing Reactor (MTR) BR2. The MYRRHA facility is currently being developed with the aid of the FP7-project “Central Design Team” and will be as a flexible irradiation facility, able to work in both subcritical and critical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications, and Si-doping. MYRRHA will also demonstrate the full concept of Accelerator Driven Systems by coupling the requisite three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow for the study of efficient transmutation of high-level nuclear waste. Since MYRRHA is based on the heavy liquid metal technology, Lead–Bismuth Eutectic, it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) technology. Further, in critical mode, MYRRHA will play the role of European Technology Pilot Plant in the path forward for LFR. In this paper we present the historical perspectives, international and high profile membership within the consortium of the MYRRHA project and the rationale for the design choices are presented. Also, the latest configuration of the reactor system is described together with the different irradiation capabilities. More specifically, the possibilities and performances for fuel irradiations are presented in detail.

  6. Breeding gains of sodium-cooled oxide-fueled fast reactors

    International Nuclear Information System (INIS)

    Mougniot, J.C.; Barre, J.Y.; Clauzon, P.; Ciacometti, C.; Neviere, G.; Ravier, J.; Sichard, B.

    Calculated values are presented for the breeding gains of French fast reactors, and the experimental uncertainties are discussed. The effect of various choices of planning on the breeding gains is next analyzed within the framework of classical concepts. In the final part, a new concept involving ''heterogeneous cores'' with a single enrichment zone is presented. This concept permits a significant improvement in the breeding gain and doubling time of fast reactors. (U.S.)

  7. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Xu Mi

    1999-01-01

    The paper outlines the recent development status of nuclear power plants in China and introduces the main design characteristics and nuclear safety features of the Chinese Experimental Fast Reactor (CEFR). During the review of the Preliminary Safety Analysis Report some important subjects have been proposed by the China National Nuclear Safety Administration (NNSA). More detailed research for the answer has been done. The main analysis results for (1) Reactor Shut-down System, (2) Decay Heat Removal System and (3) Fuel Subassembly Blockage as three examples are given in this paper. The CEFR is still in the detail design stage. Its site is almost ready for the construction of the main building. It is planned to have the first pouring of concrete in June, 1999, but it depends on the license issued by the NNSA. (author)

  8. A review of fast reactor progress in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K [Power Reactor and Nuclear Fuel Development Corporation, Tokyo (Japan)

    1978-07-01

    The fast reactor development project in Japan is continuing at a slightly increased scale of effort in budget. The total budget for LMFBR development for fiscal year 1978 was 24 billion yen. In August 1977 major industries engaged in LMFBR have set up an office where design work can be jointly conducted. Highlights and topics of the fast reactor development activities cover description of JOYO reactor, its first criticality experiment, and the prototype fast breeder MONJU. Research and development programmes dealt with fission products release and its possible interaction with the soodium coolant, inspection of reactor components, experiments simulating sodium leakage, development of steam generator.

  9. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Magny, E. de; Berte, M.

    1997-01-01

    Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

  10. Logistical and economic obstacles to a fast reactor programme

    International Nuclear Information System (INIS)

    Sweet, C.

    1982-01-01

    Fast reactor studies used to place great emphasis on its role as a breeder of plutonium, thereby raising the prospect of a nuclear power system free from the constraint of uranium supplies. Today, not only the timing of fast reactor introduction has slipped (by two or three decades) but the perspective which was central to energy policy has changed dramatically. This article first examines the fast reactor as a system and looks at the interaction of four key variables in its logistics. It then looks at the rise in real costs, especially capital costs. Given the parameters that determine the plutonium balance and the economics of the fast reactor system, the author questions whether there is a sound basis for its introduction, and concludes by suggesting that the most pressing requirement is a study of the opportunity costs of fast reactor R and D expenditures. (author)

  11. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Picker, C.; Ainsworth, K.F.

    1998-01-01

    The general position with regard to nuclear power and fast reactors in the UK during 1996 is described. The main UK Government-funded fast reactor research and development programme was concluded in 1993, to be replaced by a smaller programme which is mainly funded and managed by British Nuclear Fuels plc. The main focus of this programme sustains the UK participation in the European Fast Reactor (EFR) collaboration and the broader international links built-up over the previous decades. The status of fast reactor studies made in the UK in 1996 is outlined and, with respect to the Prototype Fast Reactor at Dounreay, a report of progress with the closure studies, fuel reprocessing and decommissioning activities is provided. (author)

  12. LTFR-4, Library Generated for Fast Reactor Design Program from JAERI Fast-Set Multigroup Constant

    International Nuclear Information System (INIS)

    Suzuki, Tomoo

    1971-01-01

    Nature of physical problem solved: The program processes JAERI-Fast group constants sets of less than 30-group and prepares a binary library tape for efficient usage by a series of related fast reactor design calculation programmes

  13. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

    Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

  14. Fast-reactor fuel reprocessing in the United Kingdom

    International Nuclear Information System (INIS)

    Allardice, R.H.; Buck, C.; Williams, J.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the United Kingdom since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium-based fast-reactor system, and the importance of establishing at an early stage fast-reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high-burnup thermal-reactor oxide fuel. The United Kingdom therefore decided to reprocess irradiated fuel from the 250MW(e) Prototype Fast Reactor (PFR) as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small-scale fully active demonstration plant has been carried out since 1972, and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste-management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant, a parallel development programme has been initiated to provide the basis for the design of a large-scale fast-reactor fuel-reprocessing plant to come into operation in the late 1980s to support the projected UK fast-reactor installation programme. The paper identifies the important differences between fast-reactor and thermal-reactor fuel-reprocessing technologies and describes some of the development work carried out in these areas for the small-scale PFR fuel-reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast-reactor fuel-reprocessing plant is outlined and the current design philosophy discussed. (author)

  15. Fast neutron nuclear reactor with lightened internal structure

    International Nuclear Information System (INIS)

    Artaud, R.; Aubert, M.; Renaux, C.

    1984-01-01

    The invention concerns an integrated type fast reactor. The inner vessel comprises two truncated shells, of which the large bases are connected either directly, or by a cylindrical shell of large diameter. The small base of the upper truncated shell is prolongated by a shell of small diameter and the small base of the lower truncated shell supports the reactor core. The invention allows the construction of simpler and less expansive fast reactors [fr

  16. Fast neutron dosimetry in research reactors

    International Nuclear Information System (INIS)

    Eckert, R.

    1960-01-01

    This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [fr

  17. Reactor physical experimental program EROS in the frame of the molten salt applying reactor concepts development

    International Nuclear Information System (INIS)

    Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

    2009-01-01

    After the relatively broad program of experimental activities, which have been involved in the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic, there has been a next stage (namely large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0 called EROS project) started, which will be focused to the experimental verification of the rector physical or neutronic properties of other types of reactor concepts applying molten salts in the role of liquid fuel and/or coolant. This tendency is based on the recently accepted decision of the MSR SSC of GIF to consider for further period of its activity two baseline concepts- fast neutron molten salt reactor non-moderated (FMSR-NM) as a long-term alternative to solid fuelled fast neutron reactors and simultaneously, advanced high temperature reactor (AHTR) with pebble bed type solid fuel cooled by liquid salts. There will be a brief description of the prepared and performed experimental programs in these directions (as well as the preliminary results obtained so far) introduced in the paper. (author)

  18. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  19. Polarization of fast neutrons in VVR-M reactor

    International Nuclear Information System (INIS)

    Garusov, E.A.; Lifshits, E.P.; Petrov, Yu.V.

    1987-01-01

    Neutron polarization in the reactor leads to circular polarization of γ quanta emitted both in radiational capture of neutrons and in the transition of nuclei excited as a result of inelastic scattering to the ground state. This may be used to determine the polarization of reactor neutrons. The circular polarization of γ quanta at light-water and graphite targets at the center of the active zone of the VVR-M reactor at the B.P. Konstantinov Leningrad Institute of Nuclear Physics was recently measured. A simplified experimental scheme is shown. Fast neutrons leaving the active zone of the reactor were excited in the inelastic scattering at the target nuclei. The polarization of the γ quanta emitted by nuclei in transitions to the ground state was measured by a polarimeter positioned above the active zone. The reason for the circular polarization of γ quanta may also be nonconservation of P parity on account of weak interaction in the capture of a neutron by hydrogen

  20. Experimental fuel channel for samples irradiation at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Markovic, H.; Sokcic-Kostic, M.; Miric, I.; Prokic, M.; Strugar, P.

    1984-12-01

    An 80% enriched UO 2 fuel channel at the RB nuclear reactor in the 'Boris Kidric' Institute of Nuclear Sciences is modified for samples irradiation by fast neutrons. Maximum sample diameter is 25 mm and length up to 1000 mm. Characteristics of neutron and gamma radiation fields of this new experimental channel are investigated. In the centre of the channel, the main contribution to the total neutron absorbed dose, i.e. 0.29 Gy/Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to intermediate and thermal neutrons, respectively. At the same time the gamma absorbed dose is 0.35 Gy. The developed experimental fuel channel, EFC, has wide possibilities for utilization, from fast neutron spectrum studies, electronic component irradiations, dosemeters testing, up to cross-section measurements. (author)

  1. Implications of Fast Reactor Transuranic Conversion Ratio

    International Nuclear Information System (INIS)

    Piet, Steven J.; Hoffman, Edward A.; Bays, Samuel E.

    2010-01-01

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 ('burners') do not have blankets; the cases above CR=1 ('breeders') have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is 'attractive' for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR 1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.

  2. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.

    1987-01-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  3. A fast spectrum dual path flow cermet reactor

    International Nuclear Information System (INIS)

    Anghaie, S.; Feller, G.J.; Peery, S.D.; Parsley, R.C.

    1993-01-01

    A cermet fueled, dual path fast reactor for space nuclear propulsion applications is conceptually designed. The reactor utilizes an outer annulus core and an inner cylindrical core with radial and axial reflector. The dual path flow minimizes the impact of power peaking near the radial reflector. Basic neutronics and core design aspects of the reactor are discussed. The dual path reactor is integrated into a 25000 lbf thrust nuclear rocket

  4. IAEA fast reactor knowledge preservation initiative. Project focus: KNK-II reactor, Karlsruhe, Germany

    International Nuclear Information System (INIS)

    2004-08-01

    This Working Material (including the attached CD-ROM) documents progress made in the IAEA's initiative to preserve knowledge in the fast reactor domain. The brochure describes briefly the context of the initiative and gives an introduction to the contents of the CD-ROM. In 2003/2004 a first focus of activity was concentrated on the preservation of knowledge related to the KNK-II experimental fast reactor in Karlsruhe, Germany. The urgency of this project was given by the impending physical destruction of the installation, including the office buildings. Important KNK-II documentation was brought to safety and preserved just in time. The CD-ROM contains the full texts of 264 technical and scientific documents describing research, development and operating experience gained with the KNK-II installation over a period of time from 1965 to 2002, extending through initial investigations, 17 years of rich operating experience, and final shutdown and decommissioning. The index to the documents on the CD-ROM is printed at the end of this booklet in chronological order and is accessible on the CD by subject index and chronological index. The CD-ROM contains in its root directory also the document 'fr c lassification.pdf' which describes the classification system used for the present collection of documents on the fast reactor KNK-II

  5. Experimental measurement of zero power reactor transfer function

    International Nuclear Information System (INIS)

    Liang Shuhong

    2011-01-01

    In order to study the zero power reactor (ZPR) transfer function, the ZPR transfer function expression was deduced with the point reactor kinetics equation, which was disturbed by reactivity input response. Based on the Fourier analysis for the input of triangular wave, the relation between the transfer function and reactivity was got. Validating research experiment was made on the DF-VI fast ZPR. After the disturbed reactivity was measured, the experimental value of the transfer function was got. According to the experimental value and the calculated value, the expression of the ZPR transfer function is proved, whereas the disturbed reactivity is got from the transfer function. (authors)

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

    International Nuclear Information System (INIS)

    Smith, R.D.

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-07-01

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

  8. Core disruptive accident analysis in prototype fast breeder reactor

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  9. Vibrations measurement in fast and PWR reactor study

    International Nuclear Information System (INIS)

    Tigeot, Y.; Epstein, A.; Hareux, F.

    1975-01-01

    In the past severe damages have occured in several nuclear reactors, by structural vibrations induced by the primary cooling flow. To avoid this kind of troubles, the SEMT makes studies for two different types of reactors. For the light pressurized water reactors, some tests have been made on the SAFRAN test loop which is a three loop 1/8 scale internal model of a 900 MWe reactor. This study is actually undertaken jointly with Framatome. Elsewhere, measurements have been made on the Phenix fast breeder sodium reactor, and studies are planned for the Super Phenix reactor [fr

  10. Status of fast reactor activities in the USSR

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Rinejskij, A.A.

    1990-01-01

    Four fast reactors are in operation in the USSR now: BR-10, BOR-60, BN-350 and BN-600. Load factor of BN-600 reactor was in 1989 about 76%. On the basis of operational experience of running reactors design of more powerful commercial size BN-800 power reactor has been completed recently and construction work has started at two sites. The BN-1600 reactor is considered to be the prototype of future commercial reactors. In 1990, it was decided to extend its design approach with the aim to find some additional solutions to provide higher safety and better economics. (author). Figs and tabs

  11. The development of fast simulation program for marine reactor parameters

    International Nuclear Information System (INIS)

    Chen Zhiyun; Hao Jianli; Chen Wenzhen

    2012-01-01

    Highlights: ► The simplified physical and mathematical models are proposed for a marine reactor system. ► A program is developed with Simulink module and Matlab file. ► The program developed has the merit of easy input preparation, output processing and fast running. ► The program can be used for the fast simulation of marine reactor parameters on the operating field. - Abstract: The fast simulation program for marine reactor parameters is developed based on the Simulink simulating software according to the characteristics of marine reactor with requirement of maneuverability and acute and fast response. The simplified core physical and thermal model, pressurizer model, steam generator model, control rod model, reactivity model and the corresponding Simulink modules are established. The whole program is developed by coupling all the Simulink modules. Two typical transient processes of marine reactor with fast load increase at low power level and load rejection at high power level are adopted to verify the program. The results are compared with those of Relap5/Mod3.2 with good consistency, and the program runs very fast. It is shown that the program is correct and suitable for the fast and accurate simulation of marine reactor parameters on the operating field, which is significant to the marine reactor safe operation.

  12. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    1985-01-01

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  13. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  14. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1995-01-01

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  15. Preparation of covariance data for the fast reactor. 2

    International Nuclear Information System (INIS)

    Shibata, Keiichi; Hasagawa, Akira

    1998-03-01

    For some isotopes important for core analysis of the fast reactor, covariance data of neutron nuclear data in the evaluated nuclear data library (JENDL-3.2) were presumed to file. Objected isotopes were 10-B, 11-B, 55-Mn, 240-Pu and 241-Pu. Physical amounts presumed on covariance were cross section, isolated and unisolated resonance parameters and first order Legendre coefficient of elastic scattering angle distribution. Presumption of the covariance was conducted in accordance with the data estimation method of JENDL-3.2 as possible. In other ward, when the estimated value was based on the experimental one, error of the experimental value was calculated, and when based on the calculated value, error of the calculated one was obtained. Their estimated results were prepared with ENDF-6 format. (G.K.)

  16. Preliminary study of a flux converter for experimental reactor

    International Nuclear Information System (INIS)

    Malouch, M.F.

    1998-01-01

    The purpose of this project is to define the characteristics of a flux converter dedicated to increase the fast neutron flux in irradiation devices placed in the core of Osiris experimental reactor. This preliminary work has dealt with the neutronic and thermal-hydraulic aspects of this problem. The synthesis of the results produced by the codes APOLLO2, DAIXY, MERCURE5.3 and FLICA-3M shows that a cylindrical converter equipped with 5 fissile rings can enhance the fast flux by a 35% factor in an experimental device set in its center. (A.C.)

  17. Theoretical and Experimental Analysis of Fast Neutron Spectra

    Energy Technology Data Exchange (ETDEWEB)

    Van Dam, H.; Kleijn, H. R. [Reactor Instituut, Delft (Netherlands)

    1968-04-15

    The reactor physics division of the Inter-Academic Reactor Institute at Delft is concentrating its efforts in the field of fast reactor physics on problems of a more fundamental nature. The object of the programme is to determine experimentally a number of microscopic reactor physics parameters such as conversion potentials, fission ratios and Doppler coefficients for simple geometries and material compositions. Because of the extreme importance of knowledge of the neutron spectrum for the interpretation of the results, attention has initially been concentrated on both the measurement and the calculation of fast neutron spectra. The transport of neutrons in absorbing and non-absorbing heavy atom materials is studied by solving the Boltzmann equation. Both isotropic and anisotropic scattering are considered. Anisotropic scattering is treated by the P{sub n}-approximation, while flux-anisotropy is handled with the S{sub N}-method. In the code FAST-DELFT, scattering is treated up to the P{sub 4} component, a further extension of which is useless because of the lack of available cross-section data. By using this method, the effect of scattering anisotropy on the spectrum formation has been studied. In addition the influence of group cross-section inaccuracies was determined. The experimental work has been concentrated on methods to determine in-core spectra. Using home-made proportional counters with gamma-ray discrimination provisions fast neutron spectra have been measured in simple geometries. These experiments were complemented by foil measurements in the lower energy region. The results of this work are presented in this paper. (author)

  18. Accuracy of helium accumulation fluence monitor for fast reactor dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Chikara; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    A helium (He) accumulation fluence monitor (HAFM) has been developed for fast reactor dosimetry. In order to evaluate the measurement accuracy of neutron fluence by the HAFM method, the HAFMs of enriched boron (B) and beryllium (Be) were irradiated in the Fast Neutron Source Reactor `YAYOI`. The number of He atoms produced in the HAFMs were measured and compared with the calculated values. As a result of this study, it was confirmed that the neutron fluence could be measured within 5 % by the HAFM method, and that met the required accuracy for fast reactor dosimetry. (author)

  19. A new safety approach in the design of fast reactors

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

    1987-01-01

    A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

  20. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

    2013-06-01

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  1. Fast reactors fuel Cycle: State in Europe

    International Nuclear Information System (INIS)

    1991-01-01

    In this SFEN day we treat all aspects (economics-reactor cores, reprocessing, experience return) of the LMFBR fuel cycle in Europe and we discuss about the development of this type of reactor (EFR project) [fr

  2. Fast reactor fuel pin behaviour modelling in the UK

    International Nuclear Information System (INIS)

    Matthews, J.R.; Hughes, H.

    1979-01-01

    Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

  3. Fast reactor fuel pin behaviour modelling in the UK

    Energy Technology Data Exchange (ETDEWEB)

    Matthews, J R [UKAEA, Harwell, Didcot, Oxon (United Kingdom); Hughes, H [Springfields Nuclear Power Development Laboratories, Springfields, Salwick, Preston (United Kingdom)

    1979-12-01

    Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

  4. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  5. Radiation environment of fusion experimental reactor

    International Nuclear Information System (INIS)

    Mori, Seiji; Seki, Yasushi

    1988-01-01

    Next step device (experimental reactor), which is planned to succeed the large plasma experimental devices such as JT-60, JET and TFTR, generates radiation (neutron + gamma ray) during its operation. Radiation (neutronic) properties of the material are basis for the study on neutron utilization (energy recovery and tritium breeding), material selection (irradiation damage and lifetime evaluation) and radiation safety (personnel exposure and radiation waste). It is necessary, therefore, to predict radiation behaviour in the reactor correctly for the engineering design of the reactor. This report describes the outline of the radiation environment of the reactor based on the information obtained by the neutronic and shielding design calculation of the fusion experimental reactor (FER). (author)

  6. Past and present role of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1978-01-01

    To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tonnes, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being commissioned and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction. Experience gained from past development work encourages further strong effort towards the commercialization of FBRs. This paper briefly reviews the important role to be played by FBRs, the development programme for FBRs, and the experience gained so far from the programme in Japan. (author)

  7. Fabrication of particulate metal fuel for fast burner reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok

    2012-01-01

    U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented

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

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

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

  9. Integrating the fuel cycle at IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1992-01-01

    During the past few years Argonne National Laboratory has been developing the Integral Fast Reactor (IFR), an advanced liquid metal reactor. Much of the IFR technology stems from Argonne National Laboratory's experience with the Experimental Breeder Reactors, EBR 1 and 2. The unique aspect of EBR 2 is its success with high-burnup metallic fuel. Irradiation tests of the new U-Pu-Zr fuel for the IFR have now reached a burnup level of 20%. The results to date have demonstrated excellent performance characteristics of the metallic fuel in both steady-state and off-normal operating conditions. EBR 2 is now fully loaded with the IFR fuel alloys and fuel performance data are being generated. In turn, metallic fuel becomes the key factor in achieving a high degree of passive safety in the IFR. These characteristics were demonstrated dramatically by two landmark tests conducted at EBR 2 in 1986: loss of flow without scram; and loss of heat sink without scram. They demonstrated that the combination of high heat conductivity of metallic fuel and thermal inertia of the large sodium pool can shut the reactor down during potentially severe accidents without depending on human intervention or the operation of active engineered components. The IFR metallic fuel is also the key factor in compact pyroprocessing. Pyroprocessing uses high temperatures, molten salt and metal solvents to process metal fuels. The result is suitable for fabrication into new fuel elements. Feasibility studies are to be conducted into the recycling of actinides from light water reactor spent fuel in the IFR using the pyroprocessing approach to extract the actinides (author)

  10. Coupled hydro-neutronic calculations for fast burst reactor accidents

    International Nuclear Information System (INIS)

    Paternoster, R.; Kimpland, R.; Jaegers, P.; McGhee, J.

    1994-01-01

    Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

  11. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Durston, J.G.

    1976-01-01

    It is stated that in a liquid metal cooled fast breeder reactor wherein the core, intermediate heat exchangers and liquid metal pumps are immersed in a pool of coolant such as Na, the intermediate heat exchangers are suspended from the roof, and ducting is provided in the form of a core tank or shroud interconnected with 'pods' housing the intermediate exchangers for directing coolant from the core over the heat exchanger tubes and thence back to the main pool of liquid metal. Seals are provided between the intermediate heat exchanger shells and the walls of their 'pods' to prevent liquid metal flow by-passing the heat exchanger tube bundles. As the heat exchangers must be withdrawable for servicing, and because linear differential thermal expansion of the heat exchanger and its 'pod' must be accommodated the seals hitherto have been of the sliding kind, generally known as 'piston ring type seals'. These present several disadvantages; for example sealing is not absolute, and the metal to metal seal gives rise to wear and fretting by rubbing and vibration. This could lead to seizure or jamming by the deposition of impurities in the coolant. Another difficulty arises in the need to accommodate lateral thermal expansion of the ducting, including the core tank and 'pods'. Hitherto some expansion has been allowed for by the use of expansible bellow pairs in the interconnections, or alternatively by allowing local deformations of the core tank 'pods'. Such bellows must be very flexible and hence constitute a weak section of the ducting, and local deformations give rise to high stress levels that could lead to premature failure. The arrangement described seeks to overcome these difficulties by use of a gas pocket trapping means to effect a seal against vertical liquid flow between the heat exchanger shell and the wall of the heat exchanger housing. Full details of the arrangement are described. (U.K.)

  12. A review of fast reactor activities in Switzerland - March 1984

    International Nuclear Information System (INIS)

    Wydler, P.

    1984-01-01

    As a result of the noncentralized government in Switzerland there is no clear national policy for the future application of nuclear energy. This is reflected in the lack of a generally agreed nuclear energy research policy in the country. Consequently, activities related to several advanced reactor concepts are funded simultaneously at similar, but relatively low levels. The total expenditure of 5.9 million Swiss Francs (approx. 1 SFr per capita) for fast reactor activities in 1983 must be judged in the light of this situation. The funds have been allocated to an LMFBR safety programme (52%) and a fuel development programme (48%). In the field of LMFBR safety analytical work is performed on hypothetical core disruptive accidents (HCDAs) and on the integrity of components under HCDA loadings with emphasis on the dynamic behaviour of the reactor cover. A considerable effort has recently been devoted to the preparations for the SONACO natural convection experiment. Another relatively new experimental activity, involving small-scale vapour explosions with freon and water, has produced evidence of interesting physical effects which are not in accord with the assumptions of current molten fuel-coolant interaction (MFCI) models. The fuel development programme has continued with the manufacture of spherepac mixed carbide fuel pins for an irradiation experiment in FFTF. However, the time scale of the experiment has suffered a set-back due to an accident in a glove box of the production line

  13. Review of fast reactor operating experience gained in 1998 in Russia. General trends of future fast reactor development

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Y.M.; Zverev, K.V.; Sarayev, O.M.; Oshkanov, N.N.; Korol'kov, A.S.

    1999-01-01

    Review of the general state of nuclear power in Russia as for 1998 is given in brief in the paper. Results of operation of BR-10, BOR-60 and BN-600 fast reactors are presented as well as of scientific and technological escort of the BN-350 reactor. The paper outlines the current status and prospects of South-Urals and Beloyarskaya power unit projects with the BN-800 reactors. The main planned development trends on fast reactors are described concerning both new projects and R and D works. (author)

  14. Radioisotopes in the primary circuit of a fast reactor

    International Nuclear Information System (INIS)

    Berlin, M.; Cauvin, M.

    1976-01-01

    In the frame of the research performed to understand the behaviour of the radioactive isotopes of iodine in the primary coolant circuit of fast reactor, a simple theoretical model is proposed. Results concerning PHENIX and RAPSODIE are given

  15. Review of fast reactor activities at OECD (NEA)

    International Nuclear Information System (INIS)

    Stephens, M.

    1981-01-01

    The Committee on the Safety of Nuclear Installations initiated several reports in 1979. Status reports are published on: the role of fission gas release in case of fuel element failure; reactivity monitoring in a LMFBR at shutdown; increasing the reliability of fast reactor shutdown systems. A report is planned on the interactions between sodium and concrete. LMFBR safety issue that were studied are concerned with containment R and D; natural circulation cooling; and fuel failure modelling. Nuclear Development Division was concerned with Gas cooled fast reactors technology. Nuclear Science Division dealt with fast reactor physics and nuclear data for fast reactors. NEA Data Bank provides technical support and acts as a computer code library and nuclear data centre

  16. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    Wheeler, R.C.; Bramman, J.I.

    1988-04-01

    The fast reactor programme in the United Kindom is reviewed under the following headings: Progress with PFR; Reprocessing: Commercial Design Studies; Structural Integrity; Engineering and Components; Materials; Sodium Chemistry; Core and Fuel; Safety; Plant Performance. (author)

  17. Fast wave current drive in reactor scale tokamaks

    International Nuclear Information System (INIS)

    Moreau, D.

    1992-01-01

    The IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks, hosted by the Commissariat a l'Energie Atomique (CEA), Departement de Recherches sur la Fusion Controlee (Centres d'Etudes de Cadarache, under the Euratom-CEA Association for fusion) aimed at discussing the physics and the efficiency of non-inductive current drive by fast waves. Relevance to reactor size tokamaks and comparison between theory and experiment were emphasized. The following topics are described in the summary report: (i) theory and modelling of radiofrequency current drive (theory, full wave modelling, ray tracing and Fokker-Planck calculations, helicity injection and ponderomotive effects, and alternative radio-frequency current drive effects), (ii) present experiments, (iii) reactor applications (reactor scenarios including fast wave current drive; and fast wave current drive antennas); (iv) discussion and summary. 32 refs

  18. Slovakia: Proposal of movable reflector for fast reactor design

    International Nuclear Information System (INIS)

    Vrban, B.

    2015-01-01

    In fast reactors a larger migration area leading to a significant leak of neutrons can be observed because especially the transport cross-sections are in general smaller as compared to light water reactors. The utilization of a moveable reflector system in conjunction with dedicated safety control rods can increase the ability of accident managing due to enhanced escaping neutrons which otherwise would be reflected back into the fuel zone. The paper demonstrates the possibility of better controlling the transient reactor by additionally moving selected reflector subassemblies equipped with the neutron trap. The main purpose of the analysis of the Gas-cooled Fast Reactor (GFR) presented in the full paper is investigation of the kinetic parameters and of the control and reflector rod worth, as well as optimization of the parts used for partial reflector withdrawal. The results found in this study may serve for future design improvements of other designs such as the liquid metal cooled fast reactors

  19. Some basic concepts of fast breeder reactor safeguards

    International Nuclear Information System (INIS)

    Tkharev, E.; Walford, F.J.

    1987-04-01

    The range of discussion topics of this report is restricted to a few key areas of safeguards importance at Fast Breeder Reactors (FBR) only. The differences between thermal and fast reactors that may have safeguards significance in the case of FBRs are listed. The FBR principles of design are mentioned. The relevant safeguards objectives and criteria are given. The fundamental issues for safeguarding FBR are treated. An outline safeguards approach is presented. Model inspection activities are mentioned. 4 figs

  20. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  1. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    1989-07-01

    The twenty-second Annual Meeting of the International Working Group on Fast Reactors took place in Vienna, 18-21 April 1989. Nineteen representatives from twelve Member States and International Organizations attended the Meeting. This publication is a collection of presentations in which the participants reported the status of their national programmes on fast breeder reactors. A separate abstract was prepared for each of the twelve papers from this collections. Refs, figs, tabs and 1 graph

  2. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-02-01

    This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

  3. Technology development of fast reactor fuel reprocessing technology in India

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2009-01-01

    India is committed to the large scale induction of fast breeder reactors beginning with the construction of 500 MWe Prototype Fast Breeder Reactor, PFBR. Closed fuel cycle is a prerequisite for the success of the fast reactors to reduce the external dependence of the fuel. In the Indian context, spent fuel reprocessing, with as low as possible out of pile fissile inventory, is another important requirement for increasing the share in power generation through nuclear route as early as possible. The development of this complex technology is being carried out in four phases, the first phase being the developmental phase, in which major R and D issues are addressed, while the second phase is the design, construction and operation of a pilot plant, called CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell. The third phase is the construction and operation of Demonstration of Fast Reactor Fuel Reprocessing Plant (DFRP) which will provide experience in fast reactor fuel reprocessing with high availability factors and plant throughput. The design, construction and operation of the commercial plant (FRP) for reprocessing of PFBR fuel is the fourth phase, which will provide the requisite confidence for the large scale induction of fast reactors

  4. Fast reactor cover gas purification - The UK position

    International Nuclear Information System (INIS)

    Thorley, A.W.

    1987-01-01

    The cover gas in the Prototype Fast Reactor (PFR) provides an inert gas blanket for both primary and secondary sodium circuits, ensures inert gas padding exists between the upper seals associated with penetrations through the reactor roof and provides argon to items of plant such as the control rods and the rotating shield and also to on line instruments such as the secondary circuit Katharometers. In order to meet these and other requirements purification of the argon cover gas is important to ensure: gas fed to purge gaps in the area of the magnetic hold device in the control rod mechanisms is not laden with sodium aerosols and reactive impurities (O 2 , H 2 ) which could cause blocking both within the gaps and pipelines; gas phase detection systems which provide early warning of steam generator failures or oil ingress into the sodium are not affected by the presence of gaseous impurities such as H 2 , CO/CO 2 and CH 4 ; mass transfer processes involving both corrosion products and interstitial atoms cannot be sustained in the cover gas environment due to the presence of high levels of O 2 , N 2 and carburising gases; background levels of radioactivity (eg Xe 133) are sufficiently low to enable gas phase detection of failed fuel pins, and the primary circuit gas blanket activity is sufficiently reduced so that discharges to the atmosphere are minimised. This paper describes how the PFR cover gas purification system is coping with these various items and how current thinking regarding the design of cover gas purification systems for a Civil Demonstration Fast Reactor (CDFR), where larger gas volumes and higher levels of radioactivity may be involved, is being guided by current experience on PFR. The paper also briefly review the experimental work planned to study aerosol and caesium behaviour in cove gas environments and discusses the behaviour of those impurities such as Zn, oil and N 2 which are potentially damaging if certain levels are exceeded in operating

  5. 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 ...

  6. Fast breeder reactors: can we learn from experience

    International Nuclear Information System (INIS)

    Keck, O.

    1981-01-01

    An economic analysis of FBRs, in particular the long-term benefits to be expected, with reference to the experience of the West German fast breeder reactor programme suggests ways of bringing more realism into governmental decisions on the development of new reactor types. It is suggested that if reactor manufacturers and utilities financed commercial-size demonstration plants from their own funds, then the government would get more realistic advice. (U.K.)

  7. Knowledge Preservation and Data Collection on Fast Reactors in France

    International Nuclear Information System (INIS)

    Vasile, A.

    2013-01-01

    Documentary packages: • MADONA - the French LMFR R&D Documentary collection; - A trilateral CEA, EDF & AREVA collection, 64 synthesis reports with 5300 relevant documents, on line. • RAPSODIE documentary collection - 18000 documents including 8000 drawings gathering available technical information on RAPSODIE; • PHENIX documentary fund - 140 000 documents. Design, specification, drawings, procedures, technical reports gathering available technical information on the PHENIX plant since 1968; • PHENIX lifetime extension project data base - 2000 reports. General methodology and feedback experience; • SUPERPHENIX Plant Archives - Several thousand of documents covering plant design, operation and decommissioning, since 1977; • SPX Safety report - 150 000 pages; - Plant characteristics & Technical options; - General safety analysis; - Manufacturing and operational rules (normal, incidental and anticipated operating conditions); • EFR 98 design documentation - around 5700 documents produced during the EFR studies; • EFR Synthesis reports - 48 synthesis reports; • ACCORE - 20 200 CEA technical documents on the period 1988-2001, with keyword research system; • The Fast Neutron Reactor System Knowledge Book - approx. 1400 licensing documents; • Sodium Technology data base - around 5500 technical documents on sodium circuits, sodium leak and fires, instrumentation, ISI&R, sodium water reaction in SG…; • CIR - The Reprocessing Information Centre data base gathered 25 000 reports, on reprocessing and analytical chemistry; among them around 600 reports concern Fast Reactor fuel reprocessing technology; • Database - Preservation of data relative to fast critical mock-ups: MASURCA (F), SNEAK(D), ZPPR (US), ZEBRA(UK), RRR/SEG(D). • ARCOPAC - Electronic data base on SUPERPHENIX physical measurements; • BREF Fuel Data Base - SFR fuels, behavior of structural materials; • Fatigue experimental data base for 316LN; • PHENIX expert Data base - Built

  8. Fast reactors: the future of nuclear energy

    International Nuclear Information System (INIS)

    Carvalho, H.G. de.

    1988-08-01

    The main problems to be solved for FBR type reactors become viable economically, presenting the research programs of Europe, United States of America, Japan and Brazil are described. The cooperations between interested countries for improving FBR type reactors, and the financial and human resources necessaries for the development of programs, are evaluated. The fuel cycle is also analysed. (M.C.K.) [pt

  9. Technical Meeting on Existing and Proposed Experimental Facilities for Fast Neutron Systems. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Existing and proposed experimental facilities for fast neutron systems” was threefold: 1) presenting and exchanging information about existing and planned experimental facilities in support of the development of innovative fast neutron systems; 2) allow creating a catalogue of existing and planned experimental facilities currently operated/developed within national or international fast reactors programmes; 3) once a clear picture of the existing experimental infrastructures is defined, new experimental facilities are discussed and proposed, on the basis of the identified R&D needs

  10. Technical Meeting on Existing and Proposed Experimental Facilities for Fast Neutron Systems. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Existing and proposed experimental facilities for fast neutron systems” is threefold: first, it is intended for presenting and exchanging information about existing and planned experimental facilities in support of the development of innovative fast neutron systems; second, it will allow to create a catalogue of existing and planned experimental facilities currently operated/developed within national or international fast reactors programmes; third, once a clear picture of the existing experimental infrastructures is defined, new experimental facilities will be discussed and proposed, on the basis of the identified R&D needs

  11. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

    2006-01-01

    One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  12. U.S. Status of Fast Reactor Research and Technology

    International Nuclear Information System (INIS)

    Hill, Robert

    2012-01-01

    Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

  13. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  14. Status of national programmes on fast reactors in Korea

    International Nuclear Information System (INIS)

    Kim, Y.I.; Hahn, D.

    2002-01-01

    The role of nuclear power plants in electricity generation in Korea is expected to become more important in the years to come due to poor natural resources and green house gases. This heavy dependence on nuclear power eventually raises the issues of efficient utilization of uranium resources and of spent fuel storage. Fast reactors can resolve these issues. Korea Atomic Energy Research Institute started development of a Liquid Metal Reactor design in 1997 and completed the Conceptual Design in March of 2002. Efforts are currently directed toward the development of advanced fast reactor concepts and basic key technologies. (author)

  15. Review of the United Kingdom fast reactor programme - March 1986

    International Nuclear Information System (INIS)

    Bramman, J.I.; John, C.T.; Wheeler, R.C.

    1986-01-01

    The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

  16. The dissolver paradox as a coupled fast-thermal reactor

    International Nuclear Information System (INIS)

    Lutz, H.F.; Webb, P.S.

    1993-05-01

    The dissolver paradox is treated as coupled fast-thermal reactors. Each reactor is sub-critical but the coupling is sufficient to form a critical system. The practical importance of the system occurs when the fast system by itself is mass limited and the thermal system by itself is volume limited. Numerous 1D calculations have been made to calculate the neutron multiplication parameters of the separate fast and thermal systems that occur in the dissolver paradox. A model has been developed to describe the coupling between the systems. Monte Carlo calculations using the MCNP code have tested the model

  17. Simulator platform for fast reactor operation and safety technology demonstration

    International Nuclear Information System (INIS)

    Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-01-01

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  18. Simulator platform for fast reactor operation and safety technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  19. A review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    Pierantoni, F.; Tavoni, R.

    1984-01-01

    This review sums up the Italian situation in the field of the fast reactors on the eve of the fifth five year plan (1985-1989), in which the country undertakes to implement an important activity of research and development in the context of a greater European collaboration. Italian participation in the development of European nuclear power stations together with the completion of the PEC plant which will be used to develop a fuel element with the necessary economic and safety characteristics, remain the two principal goals of the Italian fast reactor programme. In 1983 the sum assigned by ENEA for fast reactors was about 220 billion lire of which 145 billion was for the PEC reactor

  20. Research on the usage of a deep sea fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-09-01

    Many new types of fast reactors have been studied in PNC. A deep sea fast reactor has the highest realization probability of the reactors studied because its development is desired by many specialists of oceanography, meteorology, deep sea bottom oil field, seismology and so on and because the development does not cost big budget and few technical problems remain to be solved. This report explains the outline and the usage of the reactor of 40 kWe and 200 to 400 kWe. The reactor can be used as a power source at an unmanned base for long term climate prediction and the earth science and an oil production base in a deep sea region. On the other hand, it is used for heat and electric power supply to a laboratory in the polar region. In future, it will be used in the space. At the present time, a large FBR development plan does not proceed successfully and a realization goal time of FBR has gone later and later. We think that it is the most important to develop the reactor as fast as possible and to plant a fast reactor technique in our present society. (author)

  1. The 'Reacteur Jules Horowitz': a new experimental reactor project

    International Nuclear Information System (INIS)

    Frachet, S.; Ballagny, A.

    1999-01-01

    The Jules Horowitz Reactor (RJH) is a new research reactor project dedicated to materials and nuclear fuel testing, the location of which is foreseen at the CEA-CADARACHE site, and the start-up in 2006. The launching of this project originated from a double finding: The development of nuclear power plants aimed at satisfying the energy needs of the next century, cannot be envisaged without experimental reactors which are unrivaled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. The existing experimental reactors are 30 to 40 years old and it is advisable to examine henceforth the necessity for and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a study on the mid and long term irradiation needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfill the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The reactor project selected among several different concepts, is finally a light water pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows for many irradiations in and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and the French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the following most important items: neutronic and thermalhydraulic studies on alternative core designs, with or without added pressurization

  2. Fast Reactor Programme. Second Quarter 1969. Progress Report. RCN Report

    International Nuclear Information System (INIS)

    Hoekstra, E.K.

    1969-12-01

    This progress report covers fast reactor research carried out by RCN during the second quarter 1969 forming part of the integrated fast breeder research and development programme also in progress at the national nuclear research centres Karlsruhe and Mol. The combined effort is based on a memorandum of co-operation in this field signed by the respective governments in 1968 and on a memorandum of understanding signed by the research centres. The RCN contribution is mainly concerned with the core of the fast breeder reactor and related safety aspects and, as such, must be looked upon as being complementary to the industrial research pro- field of fast reactors. The contribution comprises the following six items: - A Æéatîtôr , physics programme to determine the influence of fission products on several main characteristics of the reactor core such as void coefficient, Doppler coefficient and breeding ratio; - A fuel performance programme in which both stationary and transient irradiations are being carried out to establish the temperature and power limits of fuel rods; also the consequences of loss- of-cooling will be investigated; - Investigation into the change in mechanical properties of fuel canning materials due to high fast neutron doses; - A study of the corrosion behaviour of canning materials and their compatibility with the fuel under conditions of high temperature and high pressure; - Investigation into the behaviour of aerosols of fission products which could be formed after a fast reactor accident; a thorough understanding is of utmost importance for the reactor safety assessment ; - Studies on heat transfer in the reactor core. As fast breeders operate at high power densities, an accurate knowledge on the heat transfer phenomena is required

  3. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1992-01-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  4. A review of fast reactor program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  5. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  6. Waste management in IFR [Integral Fast Reactor] fuel cycle

    International Nuclear Information System (INIS)

    Johnson, T.R.; Battles, J.E.

    1991-01-01

    The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

  7. The computerized reactor period measurement system for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

    The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

  8. Computer measurement system of reactor period for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

    The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

  9. 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

  10. Metallic uranium as fuel for fast reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de

    1988-01-01

    This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

  11. Fast-neutron nuclear reactor vessel

    International Nuclear Information System (INIS)

    Presciuttini, L.

    1984-01-01

    The reactor vessel comprises a cylindrical shell, of which axis is vertical, coupled at its lower part with a dished bottom. The reactor core rests on a support plate bearing on the bottom of the vessel. The above dished bottom comprises a spherical sector having the same radius and the same axis as the cylindrical shell and joining the lower part of the shell, and a spherical head of which radius is a little more important than the spherical sector one. A cylindrical support for the reactor core is attached to the vessel at the joint between the two dished sections. The invention applies more particularly to integrated type reactors cooled by liquid sodium [fr

  12. A Review of the UK Fast Reactor Programme: March 1980

    International Nuclear Information System (INIS)

    Smith, R.D.

    1980-01-01

    Towards the end of 1979 the Government announced a new programme of thermal reactor stations to be built over ten years (totalling 15GW), in addition to the two AGR stations at Torness and Heysham 'B' which had been approved by the previous Government. The first station of the new programme will be based on a Westinghouse PWR, subject to safety clearance and the outcome of a public inquiry, and it is envisaged that the remaining stations of the programme would be split between PWRs and AGRs. The AEA Chairman wrote formally to the Secretary of State for Energy in December 1979, putting forward on behalf of the Electricity Supply Authorities, NNC, BNFL and the AEA a recommended strategy for building the Commercial Demonstration Fast Reactor (CDFR), subject to normal licensing procedure and to public inquiry, so as to ensure that the key options for introducing commercial fast reactors, when required, should remain open. A Government statement is expected during the next few months. Meanwhile the level of effort on fast reactor research and development in the UK has been maintained, the fast reactor remaining the largest of the UKAEA's reactor development projects with expenditure totalling somewhat over £80M per annum. The main feature of the UK fast reactor programme has continued to be the operation of PFR (Sections 2 and 7) which is yielding a wealth of experience and of information relevant to the design of commercial fast reactors. Bum-up of standard driver fuel has reached 6-7% by heavy atoms, while specially enriched lead fuel pins have reached 11 % without failure. An extensive programme of work in the reactor and its associated steam plant was completed in March 1980 and the reactor then started its fifth power run. The fuel reprocessing plant at DNE is being commissioned and has reprocessed some of the spent fuel remaining from the DFR. It will start soon on reprocessing fuel discharged from the PFR. During the year improvements to the design of the future

  13. Opportunities for reactor scale experimental physics

    International Nuclear Information System (INIS)

    1999-01-01

    A reactor scale tokamak plasma will exhibit three areas of physics phenomenology not accessible by contemporary experimental facilities. These are: (1) instabilities generated by energetic alpha particles; (2) self-heating phenomena; and (3) reactor scale physics, which includes integration of diverse physics phenomena, each with its own scaling properties. In each area, selected examples are presented that demonstrate the importance and uniqueness of physics results from reactor scale facilities for both inductive and steady state reactor options. It is concluded that the physics learned in such investigations will be original physics not attainable with contemporary facilities. In principle, a reactor scale facility could have a good measure of flexibility to optimize the tokamak approach to magnetic fusion energy. (author)

  14. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-01-01

    Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

  15. Status of fast reactor activities in the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, M F; Rinejsjij, A A [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1992-07-01

    The power production program was developed before the disintegration of the USSR and CIS. This report covers therefore the current status of power production and consumption in in republics of the former USSR with a separate chapter on the status of nuclear power. It covers some general results concerned with fast reactors operational experience and BN-600 power plant operational experience. This includes radiological conditions at the BN-600 and reactor core operating experience. Separate chapters are devoted to BN-350, BOR-60, BR-10 and BN-800 reactors. Work devoted to large-size reactor design are described including research and development and fabrication.

  16. The safety basis of the integral fast reactor program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The Integral Fast Reactor (IFR) and metallic fuel have emerged as the US Department of Energy reference reactor concept and fuel system for the development of an advanced liquid-metal reactor. This article addresses the basic elements of the IFR reactor concept and focuses on the safety advances achieved by the IFR Program in the areas of (1) fuel performance, (2) superior local faults tolerance, (3) transient fuel performance, (4) fuel-failure mechanisms, (5) performance in anticipated transients without scram, (6) core-melt mitigation, and (7) actinide recycle

  17. Uranium utilization of light water cooled reactors and fast breeders

    International Nuclear Information System (INIS)

    Stojadinovic, Timm

    1991-08-01

    The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

  18. Status of fast reactor activities in the Russian Federation

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Rinejsjij, A.A.

    1992-01-01

    The power production program was developed before the disintegration of the USSR and CIS. This report covers therefore the current status of power production and consumption in in republics of the former USSR with a separate chapter on the status of nuclear power. It covers some general results concerned with fast reactors operational experience and BN-600 power plant operational experience. This includes radiological conditions at the BN-600 and reactor core operating experience. Separate chapters are devoted to BN-350, BOR-60, BR-10 and BN-800 reactors. Work devoted to large-size reactor design are described including research and development and fabrication

  19. Looking to the future with the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.

    1985-01-01

    During the past two years, scientists from Argonne have developed a design for an advanced breeder reactor with a closed, self-contained fuel cycle. This Integral Fast Reactor (IFR) is a pool-type, sodium-cooled reactor. It uses a new metal-alloy fuel design which overcomes the problem of swelling. The possibility of unauthorised diversion of nuclear fuel, and the need to transport plutonium to and from the site, is overcome by using a pyrometallurgical fuel reprocessing technique in a compact facility that is an integral part of the reactor plant. (author)

  20. Licensing issues for inherently safe fast reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Lee, S.; Okrent, D.

    1986-01-01

    There has been considerable interest recently in a new generation of liquid metal reactor (LMR) concepts in the US. Some significant changes in regulatory philosophy will be required if the anticipated cost advantages of inherently safe designs are to be achieved. The defense in depth philosophy will need to be significantly re-evaluated in the context of inherently safe reactors. It is the purpose of this paper to begin such a re-evaluation of this regulatory philosophy