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Sample records for lmfbr blanket research

  1. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

  2. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

    1971-01-01

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  3. Effect of reactor size on the breeding economics of LMFBR blankets

    International Nuclear Information System (INIS)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

  4. LMFBR blanket physics project progress report No. 4

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Lanning, D.D.; Kaplan, I.; Supple, A.T.

    1973-01-01

    During the period covered by the report, July 1, 1972, through June 30, 1973, work was devoted to completion of experimental measurements and data analysis on Blanket Mockup No. 3, a graphite-reflected blanket, and to initiation of experimental work on Blanket Mockup No. 4, a steel-reflected assembly designed to mock up a demonstration plant blanket. Work was also carried out on the analysis of a number of advanced blanket concepts, including the use of high-albedo reflectors, the use of thorium in place of uranium in the blanket region, and the ''parfait'' or completely internal blanket concept. Finally, methods development work was initiated to develop the capability for making gamma heating measurements in the blanket mockups. (U.S.)

  5. An experimental study of the heterogeneous LMFBR core using FCA assemblies with axial internal blanket

    International Nuclear Information System (INIS)

    Nakano, M.; Iijima, S.; Shirakata, K.; Hirota, J.

    1980-01-01

    To investigate physics properties of the heterogeneous LMFBR core and to examine the reliability of the current data and method for heterogeneous core configuration, an experimental study has been made on FCA VII-3 assemblies which have an internal blanket (IB) at midplane of the cylindrical core. Systematic experiments were carried out on the heterogeneous cores whose IBs were different in composition and thickness. A homogeneous core was also built to compare the results with those obtained on the heterogeneous cores. The sodium-void worth is not sensitive to the composition of IB. The positive void worth in the core of the 40 cm IB is lowered by about 40% compared with that in the homogeneous core. The analysis was made using the JAERI-Fast Set Version II and the diffusion code CITATION. Directional diffusion coefficients were used to take account of the axial streaming. To evaluate transport effects, the S 4 calculation was made. Comparison between the calculated and experimental results reveals the following: ksub(eff) and Pu worth in the core are not well predicted for the heterogeneous core, although they are represented satisfactorily for the homogeneous core. Reaction rates sensitive to the low-energy neutron are underestimated in the IB when they are normalized in the core. Sodium-void worths are fairly well predicted. However, the positive void worth in the heterogeneous core is underestimated, while that in the homogeneous core is overestimated. (author)

  6. MIT LMFBR blanket physics project progress report No. 7, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1976-01-01

    Work during the period was devoted primarily to a range of analytical/numerical investigations, including evaluation of means to improve external blanket designs, beneficial attributes of the use of internal blankets, improved methods for the calculation of heterogeneous self-shielding and parametric studies of calculated spectral indices. Experimental work included measurements of the ratio of U-238 captures to U-235 fissions in a standard blanket mockup, and completion of development work on the radiophotoluminescent readout of LiF thermoluminescent detectors. The most significant findings were that there is very little prospect for substantial improvement in the breeding performance of external blankets, but internal blankets continue to show promise, particularly if they are used in such a way as to increase the volume fraction of fuel inside the core envelope. An improved equivalence theorem was developed which may allow use of fast reactor methods to calculate heterogeneously self-shielded cross sections in both fast and thermal reactors

  7. Workshop on cold-blanket research

    International Nuclear Information System (INIS)

    1977-05-01

    The objective of the workshop was to identify and discuss cold-plasma blanket systems. In order to minimize the bombardment of the walls by hot neutrals the plasma should be impermeable. This requires a density edge-thickness product of nΔ > 10 15 cm -2 . An impermeable cold plasma-gas blanket surrounding a hot plasma core reduces the plasma wall/limiter interaction. Accumulation of impurities in this blanket can be expected. Fuelling from a blanket may be possible as shown by experimental results, though not fully explained by classical transport of neutrals. Refuelling of a reacting plasma had to be ensured by inward diffusion. Experimental studies of a cold impermeable plasma have been done on the tokamak-like Ringboog device. Simulation calculations for the next generation of large tokamaks using a particular transport model, indicate that the plasma edge profile can be controlled to reduce the production of sputtered impurities to an acceptable level. Impurity control requires a small fraction of the radial space to accomodate the cold-plasma layer. The problem of exhaust is, however, more complicated. If the cold-blanket scheme works as predicted in the model calculations, then α-particles generated by fusion will be transported to the cold outside layer. The Communities' experimental programme of research has been discussed in terms of the tokamaks which are available and planned. Two options present themselves for the continuation of cold-blanket research

  8. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, T.; Mukai, K.; Yamamoto, K.

    1980-01-01

    Bellows are employed as useful mechanical elements with their flexibility and imperviousness to liquid and gas in the system in which such chemically active substances as sodium are handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: control rod drive mechanism; intermediate heat exchanger; small valve; mechanical penetration assembly of the containment boundary; outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for FBR use in Japan. (author)

  9. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, Tadao; Mukai, Kazuo; Yamamoto, Ken.

    1979-11-01

    The bellows is employed as a useful mechanical element with its flexibility and imperviousness to liquid and gas in the system in which such chemically active substance as sodium is handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: - control rod drive mechanism, - intermediate heat exchanger, - small valve, - mechanical penetration assembly of the containment boundary, - outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for the FBR use in Japan. (author)

  10. Transient behaviour and inherent safety research of LMFBR power plants

    International Nuclear Information System (INIS)

    Zhu Jizhou; Wang Ping; Yu Baoan

    1995-06-01

    Fast Breeder Reactor will be the next generation reactor for nuclear electricity production, the development of FBR will give the profits of efficient utilization of nuclear resources. The fast reactor safety analysis is the foundation and key of FBR research work. Therefore, a block-oriented mathematical model for the primary system of LMFBRs was constructed, and the dynamic simulating results which have been carried out on micro-computer are presented for various transients, i.e. TOP, LOFS, LOHS. The results agree well with the corresponding results of the code NATDEMO and experiment results of EBR-II. Based on previous analysis, various methods are discussed to confirm the inherent safety of LMFBR

  11. Licensing decisions and safety research related to LMFBR accidents

    International Nuclear Information System (INIS)

    Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

    1977-01-01

    The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

  12. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

    Terai, T.; Tanaka, S.; Sze, D.-K.

    2000-01-01

    Force-free helical reactor, FFHR, is a demo-relevant heliotron-type D-T fusion reactor based on the great amount of R and D results obtained in the LHD project. Since 1993, collaboration works have made great progress in design studies of FFHR with standing on the major advantage of current-less steady operation with no dangerous plasma disruptions. There are two types of reference designs, FFHR-1 and FFHR-2, where molten Flibe (LiF-BeF2) is utilized as tritium breeder and coolant. In this paper, we present the outline of FFHR blanket design and some related R and D topics focusing on Be utilization. Beryllium is used as a neutron multiplier in the design and Be pebbles are placed in the front part of the tritium breeding zone. In a Flibe blanket, HF (TF) generated due to nuclear transmutation will be a problem because of its corrosive property. Though nickel-based alloys are thought to be intact in such a corrosive environment, FFHR blanket design does not adopt the alloys because of their induced radioactivity. The present candidate materials for the structure are low-activated ferritic steel (JLF-1), V-4Cr-4Ti, etc. They are capable to be corroded by HF in the operation condition, and Be is expected to work as a reducing agent in the system as well. Whether Be pebbles placed in a Flibe flow can work well or not is a very important matter. From this point, Be solubility in Flibe, reaction rate of the Redox reaction with TF in the liquid and on the surface of Be pebbles under irradiation, flowing behavior of Flibe through a Be pebble bed, etc. should be investigated. In 1997, in order to establish more practical and new data bases for advanced design works, we started a collaboration work of R and D on blanket engineering, where the Be research above mentioned is included. Preliminary dipping-test of Be sheets and in-situ tritium release experiment from Flibe with Be sheets have got started. (orig.)

  13. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  14. Safety research needs for carbide and nitride fueled LMFBR's. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1975-01-01

    The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

  15. Application of ultrasonic thermometry in LMFBR safety research

    International Nuclear Information System (INIS)

    Carlson, G.A.; Sullivan, W.H.; Plein, H.G.

    1977-01-01

    Ultrasonic thermometry has many potential applications in reactor safety research, where extremely high temperatures and lack of visual access may preclude the use of conventional diagnostics. An application (the in-core molten fuel pool experiment) will be described in which thoriated tungsten ultrasonic thermometers were used to measure temperatures in UO 2 to incipient melt (2860 0 ). Each thermometer included five sensor elements 10 mm long, providing five temperatures within the UO 2 at various axial locations. The 10 mm spatial resolution is about five times better than previous applications of the technique. Temperature resolution of +-10 0 C was indicated by calibration data. Besides providing temperature data approximately 1000 0 C higher than were obtained with thermocouples, the thermometer yielded valuable axial temperature profile data. Details of the sensors, exciting coils, and signal conditioning electronics will be given

  16. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  17. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    Tokuhiro, A.

    1991-11-01

    A review of recent research activities in liquid metal magnetohydrodynamics (LM-MHDs) is presented in this article. Two major reserach areas are discussed. The first topic involves the thermomechanical design issues in a proposed tokamak fusion reactor. The primary concerns are in the magneto-thermal-hydraulic performance of a self-cooled liquid metal blanket. The second topic involves the application of MHD in material processing in the metallurgical and semiconductor industries. The two representative applications are electromagnetic stirring (EMS) of continuously cast steel and the Czochralski (CZ) method of crystal growth in the presence of a magnetic field. (author) 24 figs., 10 tabs., 136 refs

  18. MIT LMFBR blanket research project. Quarterly progress report, October 1, 1980-December 31, 1980

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1981-01-01

    Information on heterogeneous assembly evaluation is presented concerning ringed moderator description and flux trap description; cross section generation; ringed moderator results; effects of heterogeneous moderation; sodium void worth breakdown; flux trap configuration; flux trap results; the effects of stoichiometric variation in the moderator; and the effects of lattice representation. Also discussed is the engineering compatibility of heterogeneous assembly designs

  19. Annual report of the CTR Blanket Engineering research facility in 1996

    International Nuclear Information System (INIS)

    1998-02-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1996. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

  20. Annual report of the CTR Blanket Engineering research facility in 1992

    International Nuclear Information System (INIS)

    1993-08-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1992. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

  1. Annual report of the CTR Blanket Engineering research facility in 1994

    International Nuclear Information System (INIS)

    1995-09-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor(CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1994. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  2. Annual report of the CTR blanket engineering research facility in 1993

    International Nuclear Information System (INIS)

    1994-08-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1993. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  3. Overview of the CTR blanket engineering research program at the University of Tokyo

    International Nuclear Information System (INIS)

    Nakazawa, Masaharu; Madarame, Haruki; Takahashi, Yoichi; Takagi, Toshiyuki

    1989-01-01

    A small overview has been given on the fusion reactor blanket engineering research program at the University of Tokyo as an introduction to the following articles, especially in its history, organization, experimental facilities and ten years research activity. (orig.)

  4. Research report on design allowable values of structural materials for LMFBR

    International Nuclear Information System (INIS)

    1978-11-01

    The present report is composed of following two main parts. i) review and re-evaluation on test results by FCI Sub-committee studies, performed from 1973 to 1976, ii) review on procedures for determining design allowable values of structural materials for LMFBR components. Re-evaluation works have been made on monotonic tensile properties at elevated temperatures, creep and creep rupture properties, creep-fatigue properties (strain rate and tensile strain hold time effects on strain fatigue properties at elevated temperatures) of Types 316 and 304 stainless steel and 2 1/4Cr-1Mo steel (base and weld metals) produced in Japan. In the first half of the present report, creep-fatigue test results obtained by FCI Sub-committee studies are subjected to re-evaluation by the present P-FCI Sub-committee. Reviews have been made on testing methods on FCI's-creep-fatigue experiments with other test data of the test materials; high temperature monotonic tensile data, creep and creep rupture data, and origin of the test materials. The data of FCI studies are compared with other reference data obtained by several Japanese laboratories. In the latter half of the present report, procedures including ASME's are reviewed for setting design allowable values for LMFBR components on the basis of high temperature strength properties obtained with materials produced in Japan. A creep rupture data of Japanese steels are issued and examined to make proposal for a design allowable stress of S sub(t) through parameter survey. (author)

  5. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  6. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  7. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    International Nuclear Information System (INIS)

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist

  8. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1979-03-01

    This document contains up-to-date data on existing or firmly decided prototype or demonstration LMFBR reactors (Table I), on planned commercial size LMFBR according to the present status of design (Table II) and on experimental fast reactors such as BOR-60, DFR, EBR-II, FERMI, FFTF, JOYO, KNK-II, PEC, RAPSODIE-FORTISSIMO (Table III). Only corrected and revised parameters submitted by the countries participating in the IWGFR are included in this document

  9. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

    International Nuclear Information System (INIS)

    Barre, B.; Camarcat, N.

    1995-01-01

    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

  10. Research program on the feasibility of pool type LMFBR in Japan

    International Nuclear Information System (INIS)

    Hattori, Sadao

    1982-01-01

    The Central Research Institute of Electric Power Industry has started the feasibility study to evaluate the possiblity of existence of large pool type FBR plants in Japan as the three-year project from fiscal 1981. The development of FBRs is indispensable for the effective use of nuclear fuel and the establishment of energy security. The knowledge on the characteristics of FBR core, sodium technology and others has advanced rapidly in Japan. At the stage of the practical reactors with large capacity, the pool type is naturally considered as the object of selection, but the aseismatic capability and safety of the large containment vessels for the pool type and the qualitative and quantitative acceptability of the research and development for the pool type are the problems. The difference between the loop type and the pool type is only the structural change arising from the difference in the arrangement of equipment. The pool type reactors have been operated already in the UK and France. The objective of the research and main subjects, the total plan and research organization, the fundamental condition of investigation, the research procedure for respective subjects, and the outline of model test are discribed. The change of design and safety standards in the future must be predicted and taken in consideration in the research. (Kako, I.)

  11. CEC activities in the field of LMFBR safety

    International Nuclear Information System (INIS)

    Balz, W.; Finzi, S.; Klersy, R.

    1976-01-01

    The aim of the ECC is to reach a common LMFBR Safety strategy in Europe. To this end the Commission promotes collaboration between the different fast reactor projects in the Community through working groups and collaborative arrangements and contributes with a research activity executed in its Joint Research Centre Ispra. A short description is given of the activity in the working groups and of the Ispra programme on LMFBR Safety. This programme covers: LMFBR thermohydraulics, fuel coolant interactions, dynamic structure loading and response, safety related material properties and whole core accident code development

  12. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

    The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

  13. Component design for LMFBR's

    International Nuclear Information System (INIS)

    Fillnow, R.H.; France, L.L.; Zerinvary, M.C.; Fox, R.O.

    1975-01-01

    Just as FFTF has prototype components to confirm their design, FFTF is serving as a prototype for the design of the commercial LMFBR's. Design and manufacture of critical components for the FFTF system have been accomplished primarily using vendors with little or no previous experience in supplying components for high temperature sodium systems. The exposure of these suppliers, and through them a multitude of subcontractors, to the requirements of this program has been a necessary and significant step in preparing American industry for the task of supplying the large mechanical components required for commercial LMFBR's

  14. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

  15. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  16. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  17. LMFBR: safety aspects

    International Nuclear Information System (INIS)

    Natta, M.

    1990-01-01

    This presentation of LMFBR safety is limited at Super Phenix reactor. After a brief description of the reactor, some details on safety systems, in normal or accidental conditions, are given. The main functions studied are: chain reaction trip, residual power evacuation, reactor containment. In heavy accident the behaviour of Super Phenix is studied which its particular characteristics and the possibilities of operators reactions. The probability of appearance and the maximum consequences of heavy accidents are given [fr

  18. Design and economic implications of heterogeneity in an LMFBR core

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1983-01-01

    Much emphasis is currently being placed in LMFBR design on reducing both the capital cost and the fuel cycle cost of an LMFBR to insure its economic competativeness without a rapid increase in the uranium prices. In this study the relationship between two core design options, their neutronic consequences, and their effect on fuel cycle cost are analyzed. The two design options are the selection of pin diameter and the degree of heterogeneity. In the case of a heterogeneous core, with a low sodium void reactivity worth this ratio of fertile internal blanket to driver assemblies is generally about 0.40. However, some advantages of cores with heterogeneity of 0.08 to 0.2 for a fixed pin diameter have been reported

  19. An electro-hydraulic servo control system research for CFETR blanket RH

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Changqi [Hefei University of Technology, Hefei 230009, Anhui (China); Tang, Hongjun, E-mail: taurustang@126.com [Hefei University of Technology, Hefei 230009, Anhui (China); Qi, Songsong [Hefei University of Technology, Hefei 230009, Anhui (China); Cheng, Yong; Feng, Hansheng; Peng, Xuebing; Song, Yuntao [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2014-11-15

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system.

  20. An electro-hydraulic servo control system research for CFETR blanket RH

    International Nuclear Information System (INIS)

    Chen, Changqi; Tang, Hongjun; Qi, Songsong; Cheng, Yong; Feng, Hansheng; Peng, Xuebing; Song, Yuntao

    2014-01-01

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system

  1. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  2. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1985-07-01

    This document has been prepared on the basis of information compiled by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains parameters of 25 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBR). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA) are presented because its design was nearly finished and most of the components were fabricated at the time when the project was terminated. Three reactors (RAPSODIE (France), DFR (UK) and EFFBR (USA)) have been shut down. However, they are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. The first LMFBRs (CLEMENTINE (USA), EBR-1 (USA), BR-2 (USSR), BR-5 (USSR)) and very special reactors (LAMPRE (USA), SEFOR (USA)) were not recommended by the members of the IWGFR to be included in the report

  3. A risk-based evaluation of LMFBR containment response under core disruptive accident conditions

    International Nuclear Information System (INIS)

    Hartung, J.; Berk, S.

    1978-01-01

    Probabilistic risk methodology is utilized to evaluate the failure modes and effects of LMFBR containment systems under Core Disruptive Accident (CDA) conditions. First, the potential causes of LMFBR containment failure under CDA conditions are discussed and categorized. Then, a simple scoping-type risk assessment of a reference design is presented to help place these potential causes of failure in perspective. The highest risk containment failure modes are identified for the reference design, and several design and research and development options which appear capable of reducing these risks are discussed. The degree to which large LMFBR containment systems must mitigate the consequences of CDA's to achieve a level of risk (for LMFBR's) comparable to the already very low risk of contemporary LWR's is explored. Based on the results of this evaluation, several suggestions are offered concerning CDA-related design goals and research and development priorities for large LMFBR's. (author)

  4. Application Effect’s Research of Vetiver Eco Blanket in Pubugou Reservoir Fluctuating Zone

    Directory of Open Access Journals (Sweden)

    Lan Huijuan

    2015-01-01

    Full Text Available To solve the ecological disasters in Pubugou Reservoir Fluctuating Zone, ecological blanket governance model is proposed in this paper, which may provide good early environment for plants’ survival in fluctuation zone, and then play the function of greening and sustainable development to ensure the slopes’ stability. Meanwhile, based on the result of vetiver ecological blanket in Hanyuan experimental zone, we find that three kinds of typical Fluctuating Zone slope’s greening effect is good, which includes the dirt piling up slope, the whole lump of rock slope and the gravel piling up slope, and it gets an average coverage of 90.3 % as well as good strength. Due to the different geological conditions, the ecological blankets’ governance effect differs from slope to slope. Using analytic hierarchy process to calculate the weight, we get the dirt piling up slope, the whole lump of rock slope and the gravel piling up slope’s weights are 0.41, 0.17, 0.42, respectively, namely, the dirt piling up slope and the gravel piling up slope have good results overall, followed by the whole lump of rock slope.

  5. Replaceable LMFBR core components

    International Nuclear Information System (INIS)

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  6. Pressure drop measurements in LMFBR wire wrapped blanket assemblies

    International Nuclear Information System (INIS)

    Chiu, C.; Hawley, J.; Rohsenow, W.M.; Todreas, N.E.

    1977-07-01

    In this experiment, measurements of subchannel static pressure for an interior and edge subchannel were taken at two elevations in two wire-wrapped 61-pin bundles. One of the bundles has geometric characteristics of P/D = 1.067 and H/D = 8.0 (4 inch lead length and 0.501 inch rod diameter) and the other bundle has geometric characteristics of P/D = 1.067 and H/D = 4.0 (2 inch lead length and 0.501 inch rod diameter). The bundle average friction factors as well as the local subchannel friction factors for both interior and edge subchannels were determined from the experimental static pressure data. The average subchannel flow rates for both edge and interior subchannels were determined in a separate experiment. Results show that two correlations suggested by Rehme and Novendstern for the bundle average friction factor cannot predict the data within the range of experimental error. The bundle average friction factors for both bundles under test were underestimated by Rehme's correlation and overestimated by Novendstern's correlation. The results of the local subchannel friction factors indicate the effect of the wire lead length is more pronounced in the interior subchannel friction factor than in the edge subchannel friction factor. As the wire wrap lead length decreases, both interior and edge subchannel friction factors increase

  7. Pressure drop measurements in LMFBR wire wrapped blanket assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chiu, C.; Hawley, J.; Rohsenow, W.M.; Todreas, N.E.

    1977-07-01

    In this experiment, measurements of subchannel static pressure for an interior and edge subchannel were taken at two elevations in two wire-wrapped 61-pin bundles. One of the bundles has geometric characteristics of P/D = 1.067 and H/D = 8.0 (4 inch lead length and 0.501 inch rod diameter) and the other bundle has geometric characteristics of P/D = 1.067 and H/D = 4.0 (2 inch lead length and 0.501 inch rod diameter). The bundle average friction factors as well as the local subchannel friction factors for both interior and edge subchannels were determined from the experimental static pressure data. The average subchannel flow rates for both edge and interior subchannels were determined in a separate experiment. Results show that two correlations suggested by Rehme and Novendstern for the bundle average friction factor cannot predict the data within the range of experimental error. The bundle average friction factors for both bundles under test were underestimated by Rehme's correlation and overestimated by Novendstern's correlation. The results of the local subchannel friction factors indicate the effect of the wire lead length is more pronounced in the interior subchannel friction factor than in the edge subchannel friction factor. As the wire wrap lead length decreases, both interior and edge subchannel friction factors increase.

  8. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  9. LMFBR plant parameters 1991

    International Nuclear Information System (INIS)

    1991-03-01

    The document has been prepared on the basis of information provided by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains updated parameters of 27 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBRs). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA), PEC (Italy), RAPSODIE (France), DFR (UK) and EFFBR (USA) are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. Two more reactors appeared in the list: European Fast Reactor (EFR) and PRISM (USA). Parameters of these reactors included in this publication are based on the data from the papers presented at the 23rd Annual Meeting of the IWGFR. All in all more than four hundred corrections and additions have been made to update the document. The report is intended for specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors

  10. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  11. Collection of summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1995

    International Nuclear Information System (INIS)

    1996-07-01

    This report meeting was held on May 22, 1995 at University of Tokyo by about 40 participants. As the topics on the fusion reactor engineering research in Japan, lectures were given on the present state and future of nuclear fusion networks and on the strong magnetic field tokamak using electromagnetic force-balanced coils being planned. Thereafter, the reports of the results of the researches which were carried out by using this experimental facility were made, centering around the subject related to the future conception 'The interface properties of fusion reactor materials and particle transport control'. The publication was made on the future conception of the basic experiment setup for fusion reactor blanket design, the application of high temperature superconductors to the advancement of nuclear fusion reactors, the modeling of the dynamic irradiation behavior of fusion reactor materials, the interface particle behavior in plasma-wall interaction, the behavior of tritium on the surface of breeding materials, and breeding materials and the behavior of tritium in plasma-wall interaction. (K.I.)

  12. Damping in LMFBR pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

    1983-06-01

    LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

  13. European research and development programme for water-cooled lithium-lead blankets: present status and future work

    International Nuclear Information System (INIS)

    Giancarli, L.; Leroy, P.; Proust, E.; Raepsaet, X.

    1992-01-01

    The European R and D programme in support of the development of water-cooled Pb-17Li blankets for DEMO aims at improving the data base concerning tritium behaviour and compatibility between blanket materials. The four main areas of the experimental programme are structural material corrosion by Pb-17Li, tritium extraction and permeation control.=, Pb-17Li physico-chemistry, and water/Pb-17Li interaction. This paper describes the most significant results obtained to date in the various experiments performed in Europe and the future programme required to complete the data base by 1994. 28 refs

  14. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme

  15. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  16. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.C.; Smith, D.L.

    1987-10-01

    The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

  17. Route survey for LMFBR spent fuel transportation analysis

    International Nuclear Information System (INIS)

    Foley, J.T.

    1977-05-01

    Descriptions are given of surveys that were made along segments of interstate highways to obtain information on objects near the right-of-ways and on highway features that constitute hazards in the event of transportation accidents. Data collected during the surveys are summarized. The work was done in support of the LMFBR Hazards Analysis which was being performed for the Division of Reactor Development and Demonstration of the U.S. Energy Research and Development Administration

  18. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1995-09-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified

  19. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1996-01-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as primary blanket materials, which have the greatest influence in determining the overall design and performance, and secondary blanket materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified. (orig.)

  20. Use of reliability in the LMFBR industry

    International Nuclear Information System (INIS)

    Penland, J.R.; Smith, A.M.; Goeser, D.K.

    1977-01-01

    This mission of a Reliability Program for an LMFBR should be to enhance the design and operational characteristics relative to safety and to plant availability. Successful accomplishment of this mission requires proper integration of several reliability engineering tasks--analysis, testing, parts controls and program controls. Such integration requires, in turn, that the program be structured, planned and managed. This paper describes the technical integration necessary and the management activities required to achieve mission success for LMFBR's

  1. Structural and containment response to LMFBR accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Fistedis, S.H.; Baker, L. Jr.; Stepnewski, D.D.; Peak, R.D.; Gluekler, E.L.

    1978-01-01

    The results of current developments in analysing the response of reactor structures and containment to LMFBR accidents are presented. The current status of analysis of the structural response of LMFBR's to core disruptive accidents, including head response, potential missile generation and the effects of internal structures are presented. The results of recent experiments to help clarify the thermal response of reactor structures to molten core debris are summarized, including the use of this data to calculate the response of the secondary containment. (author)

  2. Status of the LMFBR development

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, J J

    1975-01-01

    The development of any new power generation system which can make a major contribution to our energy needs is a multi-faceted task involving the utilization of major human and material resources. The LMFBR development, which has the potential for supplying abundant energy for generations, is therefore a large, multi-faceted program. This summary will cover (1) the need for the liquid metal fast breeder reactor, (2) an overall perspective of its development throughout the world, (3) a brief look at the in-depth technological development program in the United States, (4) a description and status of the two major projects now under way in the program, the Fast Flux Test Facility and the Clinch River Breeder Reactor Plant, and (5) a review of the plans for continued development to achieve a reliable, safe and economic power generation system for practical commercial use on the utility networks of the country.

  3. LMFBR thermal-striping evaluation

    International Nuclear Information System (INIS)

    Brunings, J.E.

    1982-10-01

    Thermal striping is defined as the fluctuating temperature field that is imposed on a structure when fluid streams at different temperatures mix in the vicinity of the structure surface. Because of the uncertainty in structural damage in LMFBR structures subject to thermal striping, EPRI has funded an effort for the Rockwell International Energy Systems Group to evaluate this problem. This interim report presents the following information: (1) a Thermal Striping Program Plan which identifies areas of analytic and experimental needs and presents a program of specific tasks to define damage experienced by ordinary materials of construction and to evaluate conservatism in the existing approach; (2) a description of the Thermal Striping Test Facility and its operation; and (3) results from the preliminary phase of testing to characterize the fluid environment to be applied in subsequent thermal striping damage experiments

  4. Welding development for LMFBR applications

    International Nuclear Information System (INIS)

    Slaughter, G.M.; Edmonds, D.P.; Goodwin, G.M.; King, J.F.; Moorhead, A.J.

    1976-01-01

    High-quality welds with suitable properties for long-time elevated-temperature nuclear service are among the most critical needs in today's welding technology. Safe, reliable, and economic generation of future power depends on welded construction in systems such as Liquid Metal Fast Breeder Reactors (LMFBRs). Rapid thermal transients in LMFBR systems at coolant temperatures around 590 to 650 0 C (1000 to 1200 0 F) could cause creep and creep-fatigue damage that is not encountered in lower temperature reactor systems. The undesirable consequences of interaction between the two working fluids - sodium and steam - in the steam generators are also of major concern. Thus sound welds that have excellent reliability over a 30-year service life are essential. Several programs are actively underway at ORNL to satisfy this critical need and selected portions of three of these programs are discussed briefly

  5. Mirror reactor blankets

    International Nuclear Information System (INIS)

    Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

    1976-01-01

    The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

  6. Blanket testing in NET

    International Nuclear Information System (INIS)

    Chazalon, M.; Daenner, W.; Libin, B.

    1989-01-01

    The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

  7. Collection of Summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1994

    International Nuclear Information System (INIS)

    1995-07-01

    The development of nuclear fusion reactors reached such stage that the generation of fusion power output comparable with the input power into core plasma is possible. At present, the engineering design of the international thermonuclear fusion experimental reactor, ITER, is advanced by the cooperation of Japan, USA, Europe and Russia, aiming at the start of operation at the beginning of 21st century. This meeting for reporting the results has been held every year, and this time, it was held on May 19, 1995 at University of Tokyo with the theme ''The interface properties of fusion reactor materials and the control of particle transport''. About 50 participants from academic, governmental and industrial circles discussed actively on the theme. Three lectures on the topics of fusion reactor engineering and materials and seven lectures on the basic experiment of fusion reactor blanket design related to the next period project were given at the meeting. (K.I.)

  8. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

    The limitations on the performance of breeding blankets in a fusion power plant are evaluated. The breeding blankets will be key components of a plant and their limitations with regard to power density, thermal efficiency and lifetime could determine to a large degree the attractiveness of a power plant. The performance of two rather well known blanket concepts under development in the frame of the European Blanket Programme is assessed and their limitations are compared with more advanced (and more speculative) concepts. An important issue is the question of which material (structure, breeder, multiplier, coatings) will limit the performance and what improvement would be possible with a 'better' structural material. This evaluation is based on the premise that the performance of the power plant will be limited by the blankets (including first wall) and not by other components, e.g. divertors, or the plasma itself. However, the justness of this premise remains to be seen. It is shown that the different blanket concepts cover a large range of allowable power densities and achievable thermal efficiencies, and it is concluded that there is a high incentive to go for better performance in spite of possibly higher blanket cost. However, such high performance blankets are usually based on materials and technologies not yet developed and there is a rather high risk that the development could fail. Therefore, it is explained that a part of the development effort should be devoted to concepts where the materials and technologies are more or less in hand in order to ensure that blankets for a DEMO reactor can be developed and tested in a given time frame. (orig.)

  9. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  10. Sodium water reaction R and D for French LMFBR

    International Nuclear Information System (INIS)

    Cambillard, E.; Finck, P.; Lapicore, A.; Simeon, C.

    1985-01-01

    This paper presents the research and development which is underway for the French LMFBR steam generator safety study. The program comprises three major areas: (1) the analysis of realistic leaks, which includes the leak evolution and its consequences; (2) the response time of leak detection systems compared to leak propagation phenomena; and (3) the guillotine rupture (DBA) studies relative to source term evaluation by experimental/calculational approach and mechanical calculations. This program has provided information for the demonstrations of the steam generator safety in respect to a sodium-water reaction

  11. Tank type LMFBR type reactors

    International Nuclear Information System (INIS)

    Shimizu, Hiroshi

    1985-01-01

    Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

  12. Structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    1983-01-01

    Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start

  13. Structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    Vaze, M.K.K.

    1983-01-01

    The use of elastic analysis for structural design of LMFBR components is discussed. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed Prototype Fast Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is same as that of Rapsodie. Nevertheless, the design had to be checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, ASME Code Section III and the Code Case N-47 are used for high temperature design. The problems faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's shakedown and plastic cycling criteria for ratchet free operation to biaxial stress fields

  14. LMFBR design and its evolution. (2) Core design of LMFBR

    International Nuclear Information System (INIS)

    Uto, Nariaki; Mizuno, Tomoyasu

    2003-01-01

    Sodium-cooled core design studies are performed. MOX fuel core with axial blanket partial elimination subassembly due to safety consideration is studied. This type of core with high internal conversion ratio possesses capability of achieving 26 months of operation cycle length and 100 GWd/t of burnup averaged over core and blanket, which are superior characteristics in view of reducing cost of power generation. Metal fuel core is also studied, and its higher breeding capability reveals a potential of better core performance such as longer operation cycle length for the same level of electricity generation, though core outlet temperature is limited to lower level due to steel cladding-metal fuel compatibility concerns. Another metal fuel core concept using single Pu enrichment and two radial regions with individual fuel pin diameters achieves 550degC of core outlet temperature identical to that of MOX fuel core, keeping operation cycle length comparable with that of MOX fuel core. This series of study results show that sodium-cooled MOX and metal fuel cores have a high flexibility in satisfying various needs including fuel cycle cost and breeding capability, depending on the stage of introducing commercialized fast reactor cycle system. (author)

  15. Status of LMFBR development project in Japan

    International Nuclear Information System (INIS)

    Nagane, G.; Akebi, M.; Matsuno, Y.

    1987-01-01

    Initiation of the LMFBR development project in Japan was decided by the Atomic Energy Commission of Japan in 1966. In 1967, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established to realize the project as a part of its tasks of a wide scope covering all the reseatch and development activities concerning fuel cycle. In the present paper the status of experimental fast reactor (Joyo), which is the first milestone of the LMFBR project, prototype fast reactor (Monju) and R and D activities supporting the project including that for larger LMFBRs in the future is described. (author)

  16. Attenuation of airborne debris from LMFBR accidents

    International Nuclear Information System (INIS)

    Morewitz, H.A.; Johnson, R.P.; Nelson, C.T.; Vaughan, E.U.; Guderjahn, C.A.; Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

    1978-01-01

    Experimental and theoretical studies have been performed to characterize the behavior of airborne particulates (aerosols) expected to be produced by hypothetical core disassembly accidents (HCDA's) in liquid metal fast breeder reactors (LMFBR's). These aerosol studies include work on aerosol transport in a 20-m high, 850-m 3 closed vessel at moderate concentrations; aerosol transport in a small vessel under conditions of high concentration (approximately 1,000 g/m 3 ), high turbulence, and high temperature (approximately 2000 0 C); and aerosol transport through various leak paths. These studies have shown that tittle, if any, airborne debris from LMFBR HCDA's would reach the atmosphere exterior to an intact reactor containment building. (author)

  17. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  18. Tritium breeding blanket

    International Nuclear Information System (INIS)

    Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

    1991-01-01

    The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

  19. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  20. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

    Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

  1. LMFBR safety. 6. Review of current issues and bibliography of literature (1977)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1978-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development. Selected bibliographic information on LMFBRs relative to the development and safety of the breeder reactor is presented for the year 1977. The bibliography consists of approximately 198 abstracts covering research and development, operating experience, and design practices. Keyword, author, and permuted-title indexes are included for completeness

  2. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-01

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  3. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  4. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  5. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  6. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

    The economic impact of employing the highly ranked blankets in the Blanket Comparison and Selection Study (BCSS) was evaluated in the context of both a tokamak and a tandem mirror power reactor (TMR). The economic evaluation criterion was determined to be the cost of electricity. The influencing factors that were considered are the direct cost of the blankets and related systems; the annual cost of blanket replacement; and the performance of the blanket, heat transfer, and energy conversion systems. The technical and cost bases for comparison were those of the STARFIRE and Mirror Advanced Reactor Study conceptual design power plants. The economic evaluation results indicated that the nitrate-salt-cooled blanket concept is an economically attractive concept for either reactor type. The water-cooled, solid breeder blanket is attractive for the tokamak and somewhat less attractive for the TMR. The helium-cooled, liquidlithium breeder blanket is the least economically desirable of higher ranked concepts. The remaining self-cooled liquid-metal and the helium-cooled blanket concepts represent moderately attractive concepts from an economic standpoint. These results are not in concert with those found in the other BCSS evaluation areas (engineering feasibility, safety, and research and development (R and D) requirements). The blankets faring well economically had generally lower cost components, lower pumping power requirements, and good power production capability. On the other hand, helium- and lithium-cooled systems were preferred from the standpoints of safety, engineering feasibility, and R and D requirements

  7. Scale modelling in LMFBR safety

    International Nuclear Information System (INIS)

    Cagliostro, D.J.; Florence, A.L.; Abrahamson, G.R.

    1979-01-01

    This paper reviews scale modelling techniques used in studying the structural response of LMFBR vessels to HCDA loads. The geometric, material, and dynamic similarity parameters are presented and identified using the methods of dimensional analysis. Complete similarity of the structural response requires that each similarity parameter be the same in the model as in the prototype. The paper then focuses on the methods, limitations, and problems of duplicating these parameters in scale models and mentions an experimental technique for verifying the scaling. Geometric similarity requires that all linear dimensions of the prototype be reduced in proportion to the ratio of a characteristic dimension of the model to that of the prototype. The overall size of the model depends on the structural detail required, the size of instrumentation, and the costs of machining and assemblying the model. Material similarity requires that the ratio of the density, bulk modulus, and constitutive relations for the structure and fluid be the same in the model as in the prototype. A practical choice of a material for the model is one with the same density and stress-strain relationship as the operating temperature. Ni-200 and water are good simulant materials for the 304 SS vessel and the liquid sodium coolant, respectively. Scaling of the strain rate sensitivity and fracture toughness of materials is very difficult, but may not be required if these effects do not influence the structural response of the reactor components. Dynamic similarity requires that the characteristic pressure of a simulant source equal that of the prototype HCDA for geometrically similar volume changes. The energy source is calibrated in the geometry and environment in which it will be used to assure that heat transfer between high temperature loading sources and the coolant simulant and that non-equilibrium effects in two-phase sources are accounted for. For the geometry and flow conitions of interest, the

  8. Operating conditions of steam generators for LMFBR's

    International Nuclear Information System (INIS)

    Ratzel, W.

    1975-01-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  9. Operating conditions of steam generators for LMFBR's

    Energy Technology Data Exchange (ETDEWEB)

    Ratzel, W

    1975-07-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  10. THE LMFBR, key to the future

    International Nuclear Information System (INIS)

    Chipman, G.L. Jr.

    1982-01-01

    This survey explains the United States prospects for utilizing the LMFBR as a mean of meeting future energy demands. Nuclear option will represent a good financial investment only when breeder will be proved as a cost-effective option. International cooperation and combined programs are very helpful to develop breeder reactor power resource

  11. LMFBR safety experiment facility planning and analysis

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Scott, J.H.

    1976-01-01

    In the past two years considerable effort has been placed on the planning and design of new facilities for the resolution of LMFBR safety issues. The paper reviews the key issues, the experiments needed to resolve them, and the design aspects of proposed new facilities. In addition, it presents a decision theory approach to selecting an optimal combination of modified and new facilities

  12. Dynamic simulation of LMFBR systems

    International Nuclear Information System (INIS)

    Agrawal, A.K.; Khatib-Rahbar, M.

    1980-01-01

    This review article focuses on the dynamic analysis of liquid-metal-cooled fast breeder reactor systems in the context of protected transients. Following a brief discussion on various design and simulation approaches, a critical review of various models for in-reactor components, intermediate heat exchangers, heat transport systems and the steam generating system is presented. A brief discussion on choice of fuels as well as core and blanket system designs is also included. Numerical considerations for obtaining system-wide steady-state and transient solutions are discussed, and examples of various system transients are presented. Another area of major interest is verification of phenomenological models. Various steps involved in the code and model verification are briefly outlined. The review concludes by posing some further areas of interest in fast reactor dynamics and safety. (author)

  13. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.; Conn, R.W.; Wolfer, W.G.; Larsen, E.N.; Sviatoslavsky, I.N.

    1978-01-01

    A novel blanket design for ICTRs is described. This blanket is used in SOLASE, the conceptual laser fusion reactor of the University of Wisconsin. The blanket to be described offers numerous advantages, including low cost, low weight, low induced radioactivity levels, the potential for hands-on maintenance, modular construction, low pressure, ability to decouple first wall and blanket coolant temperatures, adequate breeding, low tritium inventory and leakage, and sufficiently long life

  14. Wire-wrapped rod-bundle heat-transfer analysis for LMFBR

    International Nuclear Information System (INIS)

    Wong, C.N.C.; Todreas, N.E.

    1982-07-01

    Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities

  15. Pipe supports and anchors - LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.

    1983-06-01

    Pipe design and support design can not be treated as separate disciplines. A coordinated design approach is required if LMFBR pipe system adequacy is to be achieved at a reasonable cost. It is particularly important that system designers understand and consider those factors which influence support train flexibility and thus the pipe system dynamic stress levels. The system approach must not stop with the design phase but should continue thru the erection and acceptance test procedures. The factors that should be considered in the design of LMFBR pipe supports and anchors are described. The various pipe support train elements are described together with guidance on analysis, design and application aspects. Post erection acceptance and verification test procedures are then discussed

  16. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

    In the present work we consider magneto-convective flows in one of the proposed European liquid metal blankets that will be tested in the experimental fusion reactor ITER. Here the PbLi alloy is used as breeder material and helium as coolant. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma and in case of non-uniform thermal conditions. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating and cooling of walls, since only a weak forced ow is foreseen for tritium extraction in external ancillary systems. Buoyancy can therefore become very important and modify the velocity distribution and related heat transfer performance of the blanket. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. According to the last design review two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. Velocity and temperature distributions are discussed for various volumetric heat sources and magnetic field strengths.

  17. LMFBR with booster pump in pumping loop

    International Nuclear Information System (INIS)

    Rubinstein, H.J.

    1975-01-01

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

  18. Materials engineering issues, LMFBR steam generators

    International Nuclear Information System (INIS)

    Spalaris, C.N.; Challenger, K.D.; Day, R.A.; Dutina, D.; Ring, P.J.

    1976-01-01

    Selection of 2-1/4 Cr-1 Mo as the reference construction material for LMFBR steam generators assumed a balance between its known intrinsic properties and our ability to accommodate certain of its deficiencies through design allowance. A comprehensive development program was undertaken to define base data needed, confirm assumptions made relative to desired performance, minimize defects by optimization of melting, fabrication and heat treatment processes, and prepare specifications for purchasing reactor components

  19. 15 year's summary report on blanket technology and materials of mixed fuel reactor research sponsored by national '863' projects

    International Nuclear Information System (INIS)

    Xu Zengyu; Chen Jiming; Liu Xiang

    2001-01-01

    15 year's achievements of Southwestern Institute of Physics, China, in fusion technology and materials research sponsored by National '863' Engineering Projects are summarized. Many scientific and technical achievements are obtained in the researches on tritium production and recovery, doped carbon basic materials, V-alloys, 316L SS irradiation performance, B 4 C and TiC coatings, etc. Some facilities were built and some were improved for materials research. 108 references are annexed

  20. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  1. Strategies in development of advanced fuels for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo

    1976-12-01

    Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

  2. Specialists meeting on LMFBR flow induced vibrations. Summary report

    International Nuclear Information System (INIS)

    1977-12-01

    A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs

  3. Specialists meeting on LMFBR flow induced vibrations. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-01

    A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs.

  4. Issues in the selection of the LMFBR steam cycle

    International Nuclear Information System (INIS)

    Buschman, H.W.; McConnell, R.J.

    1983-01-01

    Unlike the light-water reactor, the liquid-metal fast breeder reactor (LMFBR) allows the designer considerable latitude in the selection of the steam cycle. This latitude in selection has been exercised by both foreign and domestic designers, and thus, despite the fact that over 25 LMFBR's have been built or are under construction, a consensus steam cycle has not yet evolved. This paper discusses the LMFBR steam cycles of interest to the LMFBR designer, reviews which of these cycles have been employed to date, discusses steam-cycle selection factors, discusses why a consensus has not evolved, and finally, concludes that the LMFBR steam-cycle selection is primarily one of technical philosophy with several options available

  5. Bouyancy effects on sodium coolant temperature profiles measured in an electrically heated mock-up of a 61-rod breeder reactor blanket assembly

    International Nuclear Information System (INIS)

    Engel, F.C.; Markley, R.A.; Minushkin, B.

    1978-01-01

    The paper describes test results selected to demonstrate the effect of buoyancy on the temperature profiles in a 61-rod electrically heated mock-up of an LMFBR radial blanket assembly. In these assemblies, heat transfer occurs over a wide range of complex operating conditions. The range and complexity of conditions are the result of the steep flux and power gradients which are an inherent feature of the blanket region and the power generation level in an assembly which can vary from 20 to 1100 kW

  6. Gravitational agglomeration of post-HCDA LMFBR aerosols: nonspherical particles

    International Nuclear Information System (INIS)

    Tuttle, R.F.; Loyalka, S.K.

    1982-12-01

    Aerosol behavior analysis computer programs have shown that temporal aerosol size distributions in nuclear reactor containments are sensitive to shape factors. This research investigates shape factors by a detailed theoretical analysis of hydrodynamic interactions between a nonspherical particle and a spherical particle undergoing gravitational collisions in an LMFBR environment. First, basic definitions and expressions for settling speeds and collisional efficiencies of nonspherical particles are developed. These are then related to corresponding quantities for spherical particles through shape factors. Using volume equivalent diameter as the defining length in the gravitational collision kernel, the aerodynamic shape factor, the density correction factor, and the gravitational collision shape factor, are introduced to describe the collision kernel for collisions between aerosol agglomerates. The Navier-Stokes equation in oblate spheroidal coordinates is solved to model a nonspherical particle and then the dynamic equations for two particle motions are developed. A computer program (NGCEFF) is constructed, and the dynamical equations are solved by Gear's method

  7. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

    Japan Atomic Energy Research Institute (JAERI) has been performing the development of breeding blanket for fusion power plant, as a leading institute of the development of solid breeder blankets, according to the long-term R and D program of the blanket development established by the Fusion Council of Japan in 1999. This report is an overview of development plan, achievements of element technology development and future prospect and plan of the development of the solid breeding blanket in JAERI. In this report, the mission of the blanket development activity in JAERI, key issues and roadmap of the blanket development have been clarified. Then, achievements of the element technology development were summarized and showed that the development has progressed to enter the engineering testing phase. The specific development target and plan were clarified with bright prospect. Realization of the engineering test phase R and D and completion of ITER test blanket module testing program, with universities/NIFS cooperation, are most important steps in the development of breeding blanket of fusion power demonstration plant. (author)

  8. Recent development of a CEC'S elasto-plastic-creep cyclic benchmark programme relevant to LMFBR structural integrity

    International Nuclear Information System (INIS)

    Corsi, F.; Terzaghi, A.

    1984-01-01

    It's presented the programme of elasto-plastic benchmark calculations relevant to LMFBr, which started in 1977 with the support and coordination of the Commission of the European Communities (CEC) and the participation of nuclear engineering and manufacturing companies as well as nuclear research centers of France, Germany, Italy and the United Kingdom. (E.G.) [pt

  9. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  10. Benchmark calculation programme concerning typical LMFBR structures

    International Nuclear Information System (INIS)

    Donea, J.; Ferrari, G.; Grossetie, J.C.; Terzaghi, A.

    1982-01-01

    This programme, which is part of a comprehensive activity aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, should allow to get confidence in computer codes which are supposed to provide a realistic prediction of the LMFBR component behaviour. The calculations started on static analysis of typical structures made of non linear materials stressed by cyclic loads. The fluid structure interaction analysis is also being considered. Reasons and details of the different benchmark calculations are described, results obtained are commented and future computational exercise indicated

  11. Specialists' meeting on maintenance and repair of LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    2002-01-01

    The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topic areas were discussed by participants: National review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; Research and Development work on maintenance and repair; Experience on steam generator maintenance and repair. During the meeting papers were presented by the participants on behalf of their countries and organizations. A final discussion session was held and summaries, general conclusions and recommendations were approved by consensus

  12. Prediction of flow recirculation in a blanket assembly under worst-case natural-convection conditions

    International Nuclear Information System (INIS)

    Khan, E.U.; Rector, D.R.

    1982-01-01

    Reactor fuel and blanket assemblies within a Liquid Metal Fast Breeder Reactor (LMFBR) can be subjected to severe radial heat flux gradients. At low-flow conditions, with power-to-flow ratios of nearly the same magnitude as design conditions, buoyancy forces cause flow redistribution to the side of a bundle with the higher heat generation rate. Recirculation of fluid within a rod bundle can occur during a natural convection transient because of the combined effect of flow coastdown and buoyancy-induced redistribution. An important concern is whether recirculation leads to high coolant temperatures. For this reason, the COBRA-WC code was developed with the capability of modeling recirculating flows. Experiments have been conducted in a 2 x 6 rod bundle for flow and power transients to study recirculation in the mixed-convection (forced cooled) and natural-convection regimes. The data base developed was used to validate the recirculation module in the COBRA-WC code. COBRA-WC code calculations were made to predict flow and temperature distributions in a typical LMFBR blanket assembly for the worst-case, natural-circulation transient

  13. LMFBR accident delineation study: approach and preliminary results

    International Nuclear Information System (INIS)

    Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

    1979-01-01

    Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

  14. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Aubert, M.; Mathieu, B.; Petrequin, P.

    1983-06-01

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  15. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Uda, Tatsuhiko; Maki, Koichi.

    1993-01-01

    The present invention provides a blanket of a thermonuclear device which produces tritium fuels consumed in plasmas while converting neutrons generated in the plasmas into heat energy. That is, zirconium is coated to at least one of neutron breeder pebbles and breeder pebbles, to suppress reaction between them by being in direct contact with each other at a high temperature. Further, fins are attached to a cooling pipe at a pitch smaller than the diameter of both of the pebbles, to prevent direct contact at whole surface of the pebbles and the cooling pipe, which would lower a temperature excessively. The length of the fin is controlled to control the thickness of a helium gas gap. With such constitution, direct contact of neutron breeder pebbles and the breeder pebble which are to be filled and mixed, and tend to react at a high temperature, can be prevented. The temperature of the breeding blanket is reliably prevented from lowering below a tritium emitting temperature. The structure is simplified and the production is facilitated. (I.S.)

  16. Applicability of the Reactor Safety Study (WASH-1400) to LMFBR risk assessments

    International Nuclear Information System (INIS)

    El-Sheikh, K.A.; Feller, K.G.; Fleischer, L.; Greebler, P.; McDonald, A.; Sultan, P.; Temme, M.I.; Fullwood, R.R.

    1976-01-01

    The feasibility of applying the WASH-1400 methods and data to LMFBR risk assessment is evaluated using the following approach for a selected LMFBR: (1) Structuring the LMFBR risk assessment problem in a modular form similar to WASH-1400; (2) Comparing the predictive tools applicable to each module; (3) Comparing the dependencies among the various modules. It is concluded that the WASH-1400 applicability is limited due to LWR-LMFBR differences in operating environments and accident phenomena. WASH-1400 and LMFBR specific methods applicable to LMFBR risk assessments are indicated

  17. LMFBR subassembly response to local pressure loadings: an experimental approach

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1975-01-01

    An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

  18. Study on the temperature control mechanism of the tritium breeding blanket for CFETR

    Science.gov (United States)

    Liu, Changle; Qiu, Yang; Zhang, Jie; Zhang, Jianzhong; Li, Lei; Yao, Damao; Li, Guoqiang; Gao, Xiang; Wu, Songtao; Wan, Yuanxi

    2017-12-01

    The Chinese fusion engineering testing reactor (CFETR) will demonstrate tritium self- sufficiency using a tritium breeding blanket for the tritium fuel cycle. The temperature control mechanism (TCM) involves the tritium production of the breeding blanket and has an impact on tritium self-sufficiency. In this letter, the CFETR tritium target is addressed according to its missions. TCM research on the neutronics and thermal hydraulics issues for the CFETR blanket is presented. The key concerns regarding the blanket design for tritium production under temperature field control are depicted. A systematic theory on the TCM is established based on a multiplier blanket model. In particular, a closed-loop method is developed for the mechanism with universal function solutions, which is employed in the CFETR blanket design activity for tritium production. A tritium accumulation phenomenon is found close to the coolant in the blanket interior, which has a very important impact on current blanket concepts using water coolant inside the blanket. In addition, an optimal tritium breeding ratio (TBR) method based on the TCM is proposed, combined with thermal hydraulics and finite element technology. Meanwhile, the energy gain factor is adopted to estimate neutron heat deposition, which is a key parameter relating to the blanket TBR calculations, considering the structural factors. This work will benefit breeding blanket engineering for the CFETR reactor in the future.

  19. LMFBR safety and sodium boiling

    Energy Technology Data Exchange (ETDEWEB)

    Hinkle, W.D.; Tschamper, P.M.; Fontana, M.H.; Henry, R.E.; Padilla, A. Jr.

    1978-01-01

    Within the U.S. Fast Breeder Reactor Safety R and D Work Breakdown Structure for Line of Assurance 2, Limit Core Damage, the influence of sodium boiling upon the progression and termination of accidents is being studied in loss of flow, transient overpower, loss of piping integrity, loss of shutdown heat removal system and local fault situations. The pertinent analytical and experimental results of this research to date are surveyed and compared with the requirements for demonstrating the effectiveness of this line of assurance. A discussion of specific technical issues concerned with sodium boiling and the need for future development work is also presented.

  20. Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Allard, L.T.; Elerath, J.G.

    1976-02-01

    This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

  1. CRAB-II: a computer program to predict hydraulics and scram dynamics of LMFBR control assemblies and its validation

    International Nuclear Information System (INIS)

    Carelli, M.D.; Baker, L.A.; Willis, J.M.; Engel, F.C.; Nee, D.Y.

    1982-01-01

    This paper presents an analytical method, the computer code CRAB-II, which calculates the hydraulics and scram dynamics of LMFBR control assemblies of the rod bundle type and its validation against prototypic data obtained for the Clinch River Breeder Reactor (CRBR) primary control assemblies. The physical-mathematical model of the code is presented, followed by a description of the testing of prototypic CRBR control assemblies in water and sodium to characterize, respectively, their hydraulic and scram dynamics behavior. Comparison of code predictions against the experimental data are presened in detail; excellent agreement was found. Also reported are experimental data and empirical correlations for the friction factor of the absorber bundle in the entire flow range (laminar to turbulent) which represent an extension of the state-of-the-art, since only fuel and blanket assemblies friction factor correlations were previously reported in the open literature

  2. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  3. Status of U.S. LMFBR programme

    International Nuclear Information System (INIS)

    Yevich, J.

    1978-01-01

    The determents of the decision for deterrence of commercial reprocessing and further demonstration of the plutonium breeder were based on two premises: time is needed to establish the programme for non-proliferating fuel cycle and there is a lessened sense of urgency for the USA to establish a commercial breeder in the near future. A strong, well funded base technology effort remains and will continue until institutional and technical solutions can be found to minimize or eliminate the proliferation risk. An LMFBR option will be maintained. The FFTF will be coming on line providing a powerful tool in breeder fuel and materials development and a baseline from which to scale up heat transfer systems and components. Sodium system hardware development and testing will continue to have high priority

  4. Work plan: transient release from LMFBR fuel

    International Nuclear Information System (INIS)

    Kress, T.S.; Parker, G.W.; Fontana, M.H.

    1975-09-01

    The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

  5. Acoustic leak detection of LMFBR steam generator

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo

    1993-01-01

    The development of a water leak detector with short response time for LMFBR steam generators is required to prevent the failure propagation caused by the sodium-water reaction and to maintain structural safety in steam generators. The development of an acoustic leak detector assuring short response time has attracted. The purpose of this paper is to confirm the basic detection feasibility of the active acoustic leak detector, and to investigate the leak detection method by erasing the background noise by spectrum analysis of the passive acoustic leak detector. From a comparison of the leak detection sensitivity of the active and the passive method, the active method is not influenced remarkably by the background noise, and it has possibility to detect microleakage with short response time. We anticipate a practical application of the active method in the future. (author)

  6. Nuclear welding, application for an LMFBR

    International Nuclear Information System (INIS)

    Patriarca, P.; Goodwin, G.M.

    1975-01-01

    Fabrication of an LMFBR system is discussed, with emphasis on areas where joint welding innovations have been introduced. Each major component of the system, including reactor vessel, intermediate heat exchanger, steam generator, and sodium-containment piping, is treated separately. Developmet of special filler metals to avoid the low elevated-temperature creep ductility obtained with conventional austenitic stainless steel weldments is reported. Bore-side welding of steam generator tube-to-tubesheet joints with and without filler metal is desirable to improve inspectability and eliminate the crevice inherent with face-side weld design, thus minimizing corrosion problems. Automated welding methods for sodium-containment piping are summarized which iminimize and control distortion and ensure welds of high integrity. Selection of materials for the various components is discussed for plants presently under construction, and materials predictions are made for future concepts. (U.S.)

  7. Microprocessor-based integrated LMFBR core surveillance

    International Nuclear Information System (INIS)

    Gmeiner, L.

    1984-06-01

    This report results from a joint study of KfK and INTERATOM. The aim of this study is to explore the advantages of microprocessors and microelectronics for a more sophisticated core surveillance, which is based on the integration of separate surveillance techniques. Due to new developments in microelectronics and related software an approach to LMFBR core surveillance can be conceived that combines a number of measurements into a more intelligent decision-making data processing system. The following techniques are considered to contribute essentially to an integrated core surveillance system: - subassembly state and thermal hydraulics performance monitoring, - temperature noise analysis, - acoustic core surveillance, - failure characterization and failure prediction based on DND- and cover gas signals, and - flux tilting techniques. Starting from a description of these techniques it is shown that by combination and correlation of these individual techniques a higher degree of cost-effectiveness, reliability and accuracy can be achieved. (orig./GL) [de

  8. Low cycle fatigue of irradiated LMFBR materials

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1976-01-01

    A review of low cycle fatigue data on irradiated LMFBR materials was conducted and extensive graphical representations of available data are presented. Representative postirradiation tensile properties of annealed 304 and 316 SS are selected and employed in several predictive methods to estimate irradiated material fatigue curves. Experimental fatigue data confirm the use of predictive methods for establishing conservative design curves over the range of service conditions relevant to such CRBRP components as core former, fixed radial shielding, core barrel, lower inlet module and upper internals structures. New experimental data on fatigue curves and creep-fatigue interaction in irradiated 20 percent cold worked (CW) 316 SS and Alloy 718 would support the design of removable radial shielding and upper internals in CRBRP. New experimental information on notched fatigue behavior and cyclic stress-strain curves of all these materials in the irradiated condition could provide significant design data

  9. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies

  10. 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

  11. Maintenance and repair of LMFBR steam generators: specialists` meeting, O-Arai Engineering Center, Japan, 4-8 June 1984. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The Specialists` Meeting on "Maintenance and Repair of LMFBR Steam Generators" was held in Oarai, Japan, from 4-8 June 1984. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by the Power Reactor and Nuclear Fuel Development Corporation of Japan. The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topical areas were discussed by participants: national review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; research and development work on maintenance and repair; and experience on steam generator maintenance and repair.

  12. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

    Enoeda, M.; Sato, S.; Hatano, T.

    2001-01-01

    Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

  13. Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

    Science.gov (United States)

    Leshukov, A. Yu.; Strebkov, Yu. S.; Sviridenko, M. N.; Safronov, V. M.; Putrik, A. B.

    2017-12-01

    In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 "Blanket First Wall" of February 14, 2014, and 1.6.P3.RF.01 "Blanket Module Connections" of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m2. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015-2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

  14. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

    In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

  15. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    Smith, D.L.; Sze, D.K.

    1986-02-01

    The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

  16. Comments on US LMFBR steam generator base technology

    International Nuclear Information System (INIS)

    Simmons, W.R.

    1984-01-01

    The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects

  17. Hydrogen formation and control under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Armstrong, G.R.; Wierman, R.W.

    1976-09-01

    The objective of this study is to experimentally investigate the potential for autoignition and combustion of hydrogen-sodium mixtures which may be produced in LMFBR accidents. The purpose and ultimate usefulness of this work is to provide data that will establish the validity and acceptability of mechanisms inherent to the LMFBR that could either prevent or delay the accumulation of hydrogen gas to less than 4 percent (V) in the Reactor Containment Building (RCB) under accident conditions. The results to date indicate that sodium and sodium-hydrogen mixtures such as may be expected during LMFBR postulated accidents will ignite upon entering an air atmosphere and that the hydrogen present will be essentially all consumed until such time that the oxygen concentration is depleted

  18. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

    Sze, D.K.; Jung, J.; Cheng, E.T.

    1986-01-01

    Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

  19. The blanket interface to TSTA

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Grimm, T.L.; Sze, D.K.; Anderson, J.L.; Bartlit, J.R.; Naruse, Y.; Yoshida, H.

    1988-01-01

    The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

  20. Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

    International Nuclear Information System (INIS)

    Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

    1988-01-01

    This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

  1. Validation study of the COBRA-WC computer program for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Khan, E.U.; Bates, J.M.

    1982-01-01

    The COBRA-WC (Whole Core) computer program has been developed as a benchmark code to predict flow and temperature fields in LMFBR rod bundles. Consequently, an extensive validation study has been conducted to reinforce its credibility. A set of generalized parameters predicts data well for a wide range of geometries and operating conditions which include conventional (current generation LMFBRs) fuel and blanket assembly geometry in the forced, mixed, and natural convection regimes. The data base used for validating COBRA-WC was obtained from out-of-pile and in-pile tests. Most of the data was obtained in fully heated bundles with bundle power skew across flats up to 3:1 (max:min), Reynolds number between 500 and 80,000, and coolant mixed-mean temperature rise (δ anti T) in the range, 78 0 F less than or equal to δ anti T less than or equal to 340 0 F. Within the bundle, 95% of the predicted coolant temperature data points fall within +-25 0 F for 150 0 F less than or equal to δ anti T less than or equal to 340 0 F and within +-17 0 F for 78 0 F less than or equal to δ anti T less than or equal to 150 0 F

  2. Simulation of LMFBR pump transients and comparison to LOF that occurred at EBR-II

    International Nuclear Information System (INIS)

    Koenig, F.F.; Dean, E.M.

    1985-01-01

    In a large LMFBR plant design, a number of pumps in parallel will feed the core. It must be demonstrated that the plant can continue to operate with the loss of one of the primary pumps. It is desirable not to have check valves in the loop from a reliability and economic standpoint. Simulations have been made to determine the consequences of a loss of one pump in a four-loop pool plant in which no plant protection action is taken. This analysis would be used to determine the required power rundown that would accompany pump loss. The two primary centrifugal pumps in EBR-II feed the core and blanket plenums in two parallel flow paths. The loss of one pump will result in decrease core flow and reverse flow through the down pump since no check valves are present in the system. For a large pool plant with four primary pumps, the loss of one pump will also result in reverse flow through the down pump if check valves of flow diodes are not included. The resulting flow transient has been modeled for EBR-II and the large plant using the DNSP program

  3. Blanket maintenance by remote means using the cassette blanket approach

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

    Induced radioactivity in the blanket and other parts of a fusion reactor close to the plasma zone will dictate remote assembly, disassembly, and maintenance procedures. Time will be of the essence in these procedures. They must be practicable and certain. This paper discusses the reduction of a complicated Tokamak reactor to a simpler assembly via the use of a vacuum building in which to house the reactor and the introduction in this new model of cassette blanket modules. The cassettes significantly simplify remote handling

  4. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-02-24

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included.

  5. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  6. Bubble behavior in LMFBR core disruptive accidents. Annual report, June 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Kennedy, M.F.; Rao, S.P.; Refling, J.G.

    1976-08-01

    The work reported here is part of the Aerosol Release and Transport program for LMFBR safety assessment for the Reactor Safety Research Division of the U.S. Nuclear Regulatory Commission. Six areas were at various stages of investigation during this reporting period. A study of nonequilibrium mass transfer during fuel expansion and a study of the dynamics of fuel expansion into the sodium pool were completed. Studies are underway on condensation on above-core structures and on generation of aerosols from condensation. Studies were initiated on small-particle generation from hydrodynamic fragmentation, on particle kinematics and on particle-surface interaction

  7. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-03-21

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  8. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  9. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included

  10. LMFBR safety. 5. Review of current issues and bibliography of literature (1975--1976)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    The current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA), are discussed. Bibliographic information on worldwide LMFBRs relative to the development and safety of the breeder reactor is presented for the period 1975 through 1976. The bibliography consists of approximately 1618 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Keyword, author, and permuted-title indexes are included for completeness

  11. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-11-22

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  12. LMFBR safety. 5. Review of current issues and bibliography of literature (1975--1976)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-06-08

    The current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA), are discussed. Bibliographic information on worldwide LMFBRs relative to the development and safety of the breeder reactor is presented for the period 1975 through 1976. The bibliography consists of approximately 1618 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Keyword, author, and permuted-title indexes are included for completeness.

  13. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-08-16

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  14. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  15. Upper shielding body in LMFBR type reactors

    International Nuclear Information System (INIS)

    Shoji, Koichi.

    1986-01-01

    Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

  16. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  17. Computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

    1977-01-01

    Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

  18. Coolant mixing in the LMFBR outlet plenum

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-06-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

  19. Review of PRA methodology for LMFBR

    International Nuclear Information System (INIS)

    Yang, J. E.

    1999-02-01

    Probabilistic Risk Assessment (PRA) has been widely used as a tool to evaluate the safety of NPPs (Nuclear Power Plants), which are in the design stage as well as in operation. Recently, PRA becomes one of the licensing requirements for many existing and new NPPs. KALIMER is a Liquid Metal Fast Breeder Reactor (LMFBR) being developed by KAERI. Since the design concept of KALIMER is similar to that of the PRISM plant developed by GE, it would be appropriate to review the PRA methodology of PRISM as the first step of KALIMER PRA. Hence, in this report summarizes the PRA methodology of PRISM plant, and the required works for the PSA of KALIMER based on the reviewed results. The PRA technology of PRISM plant consists of following five major tasks: (1) development of initiating event list, (2) development of system event tree, (3) development of core response event tree, (4) development of containment response event tree, and (5) consequences and risk estimation. The estimated individual and societal risk measures show that the risk from a PRISM module is substantially less than the NRC goal. Each task is compared to the PRA methodology of Light Water Reactor (LWR)/Pressurized Heavy Water Reactor (PHWR). In the report, each task of PRISM PRA methodology is reviewed and compared to the corresponding part of LWR/PHWR PSA performed in Korea. The parts that are not modeled appropriately in PRISM PRA are identified, and the recommendations for KALIMER PRA are stated. (author). 14 refs., 9 tabs., 4 figs

  20. Intelligent type sodium instrumentations for LMFBR

    International Nuclear Information System (INIS)

    Chen Daolong

    1996-07-01

    The constructions and performances of lots of newly developed intelligent type sodium instrumentations are described. The graduation characteristic equations for corresponding transducer using the medium temperature as the parameter are given. These intelligent type sodium instrumentations are possessed of good linearity. The accurate measurement data of sodium process parameters (flowrate, pressure and level) can be obtained by means of their on-line compensation function of the temperature effect. Moreover, these intelligent type sodium instrumentations are possessed of the self-inspection, the electric shutoff protection, the setting of full-scale, the setting of alarm limits (two upper limits and two lower limits alarms), the thermocouple breaking alarm, mutual isolative the 0∼10 V direct-current analogue output and the CENTRONICS standard digital output, and the alarm relay contact output. Theses intelligent type sodium instrumentations are suitable particularly for the instrument, control and protective systems of LMFBR by means of these excellent functions based on microprocessor. The basic errors of the intelligent type sodium flowmeter, immersed sodium flowmeter, sodium manometer and sodium level gauge are +-2%, +-2.3%, +-0.3% and +-1.9% of measuring ranges respectively. (9 figs.)

  1. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  2. Cesium vapor cycle for an advanced LMFBR

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

    A review indicates that a cesium vapor topping cycle appears attractive for use in the intermediate fluid circuit of an advanced LMFBR designed for a reactor outlet temperature of 1250 0 F or more and would have the following advantages: (1) it would increase the thermal efficiency by about 5 to 10 points (from approximately 40 percent to approximately 45 to 50 percent) thus reducing the amount of waste heat rejected to the environment by 15 to 30 percent. (2) the higher thermal efficiency should reduce the overall capital cost of the reactor plant in dollars per kilowatt. (3) the cesium can be distilled out of the intermediate fluid circuit to leave it bone-dry, thus greatly reducing the time and cost of maintenance work (particularly for the steam generator). (4) the large volume and low pressure of the cesium vapor region in the cesium condenser-steam generator greatly reduces the magnitude of pressure fluctuations that might occur in the event of a leak in a steam generator tube, and the characteristics inherent in a condenser make it easy to design for rapid concentration of any noncondensibles that may form as a consequence of a steam leak into the cesium region so that a steam leak can be detected easily in the very early stages of its development

  3. 54Mn release from LMFBR cores

    International Nuclear Information System (INIS)

    Polley, M.V.

    1976-10-01

    The inventory of 54 Mn per unit exposed area of stainless steel in LMFBR cores may be calculated using a formula originally derived at HEDL. This treats the simultaneous production by activation and release by corrosion and diffusion of 54 Mn and assumes that the concentration at the steel surface is zero. The inventory per unit exposed area is calculated as a function of temperature and is compared with that calculated simply by assuming stoichiometric corrosion. An effective diffusion coefficient is used in the calculations which include contributions from both lattice and grain boundary diffusion. A general relationship is derived for the effective diffusion coefficient and it is shown how values may be obtained using the Levine-MacCallum and the Fisher theories of grain boundary diffusion. Values of the lattice diffusion coefficient were obtained by analysing data obtained from sodium loop experiments. The effect on the inventory due to the possible formation of a ferrite layers on the exposed surface is discussed and it is also shown how the inventory over several fuel cycles may be calculated. (U.K.)

  4. Measurements of dynamic shape factors of LMFBR aggregate aerosols

    International Nuclear Information System (INIS)

    Allen, M.D.; Moss, O.R.; Briant, J.K.

    1980-01-01

    Dynamic shape factors for branched, chain-like aggregates of LMFBR mixed-oxide fuels have been measured with a LAPS spiral-duct centrifuge. The aerosol was generated by repeatedly pulsing a focused laser beam onto the surface of a typical LMFBR fuel pellet. The measured values of the dynamic shape factor, corrected for slip, vary between kappa = 3.60 at D/sub ae/ = 0.5 μm, and kappa = 2.23 at D/sub ae/ = 1.5 μm

  5. Safety consequences of local initiating events in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.

  6. Airborne effluent control for LMFBR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-01-01

    A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including 131 I, krypton, tritium, 129 I, and most recently 14 C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems

  7. Safety consequences of local initiating events in an LMFBR

    International Nuclear Information System (INIS)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered

  8. Applications of simulation experiments in LMFBR core materials technology

    International Nuclear Information System (INIS)

    Appleby, W.K.

    1976-01-01

    The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

  9. Status of gamma-ray heating characterization in LMFBR

    International Nuclear Information System (INIS)

    Gold, R.

    1975-11-01

    Efforts to define gamma-ray heating in Liquid Metal Fast Breeder Reactor (LMFBR) environments have been surveyed. Emphasis is placed on both current practice for the Experimental Breeder Reactor-II (EBR-II) and future needs of the Fast Flux Test Facility (FFTF). Experimental and theoretical work are included in this preliminary survey for both high and low power environments. Current ''state-of-the-art'' accuracies and limitations are assessed. On this basis, it is concluded that a broad and sustained effort be initiated to meet requested FFTF goal accuracies. To this end, recommendations are advanced for improving the current status of gamma heating characterization and temperature measurements in LMFBR

  10. Neutronic characteristics simulation of LMFBR of great size

    International Nuclear Information System (INIS)

    Kim, Y.C.

    1987-09-01

    The CONRAD experimental program to be executed on the critical mockup MASURCA in Cadarache and use all the european plutonium stock. The objectives of this program are to reduce the uncertainties on important project parameters such as the reactivity value of control rods, the flux distribution to valid calcul methods and data to use for new LMFBR conception (heterogeneous axial core by example) and to resolve the neutronic control problems for a LMFBR of great size. The present study has permitted to define this program and its physical characteristics [fr

  11. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

    General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

  12. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou

    1998-01-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  13. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  14. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

    The fusion blanket surrounds the burning hydrogen core of the fusion reactor. It is in this blanket that most of the energy released by the DT fusion reaction is converted into useable product, and where tritium fuel is produced to enable further operation of the reactor. Blankets will involve new materials, conditions and processes. Several recent fusion blanket concepts are presented to illustrate the range of ideas

  15. Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

    International Nuclear Information System (INIS)

    Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

    2006-01-01

    In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

  16. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

    Billone, M.C.; Dienst, W.; Noda, K.; Roux, N.

    1993-11-01

    The databases for solid breeder ceramics (Li 2 ,O, Li 4 SiO 4 , Li 2 ZrO 3 and LiAlO 2 ) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized

  17. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

    Materials with structures suitable for this purpose are studied. A bibliographic review of the main solid and liquid lithiated compounds is then presented. Erosion, dimensioning and maintenance problems associated with the limiter and the first wall of the reactor are studied from the point of view of the constraints they impose on the design of the blankets. Detailed studies of the main solid and liquid blanket concepts enable the best technological compromises to be determined for the indispensable functions of the blanket to be assured under acceptable conditions. Our analysis leads to four classes of solution, which cannot at this stage be considered as final recommendations, but which indicate what sort of solutions it is worthwhile exploring and comparing in order to be in a position to suggest a realistic blanket at the time when plasma control is sufficiently good for power reactors to be envisaged. Some considerations on the general architecture of the reactor are indicated. Energy storage with pulsed reactors is discussed in the appendix, and a first approach made to minimizing the total tritium recovery [fr

  18. Feasibility study for adapting ITREC plant to reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Moccia, A.; Rolandi, G.

    1976-05-01

    The report evaluates the feasibility of adapting ITREC plant to the reprocessing LMFBR fuels, with the double purpose of: 1) recovering valuable Pu contained in these fuels and recycling it to the fabrication plant; 2) trying, on a pilot scale, the chemical process technology to be applied in a future industrial plant for reprocessing the fuel elements discharged from fast breeder power reactors

  19. Refractory metal carbide coatings for LMFBR applications: a systems approach

    International Nuclear Information System (INIS)

    Gotschall, H.L.; Ople, F.S.; Riccardella, P.C.

    1975-01-01

    The selection, testing and improvement of high density, tightly bonded plasma and detonation gun coatings designed to meet LMFBR core component criteria are described. The process descriptions include a review of the important developments in substrate surface preparation which were required to ensure strong bonding and to minimize interface contamination. Coating finishing techniques which were developed to optimize friction behavior are also described

  20. Technical considerations relative to removal of sodium from LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, J S; Asquith, J G

    1975-07-01

    Reviewed in this paper are technical considerations which are of importance in choosing between an alcohol process and a moist nitrogen process for the removal of sodium from LMFBR components. Results observed in laboratory tests and in the cleaning of large scale components (e.g. a 28 MWt Modular Steam Generator Test Unit) are presented and discussed. (author)

  1. German position paper on structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    Angerbauer, A.; Link, F.

    1983-01-01

    During the design period of the German LMFBR, the SNR-300, extensive work had been done in the field of elastic and inelastic analysis. Furthermore, special design rules have been developed. A review of these activities and their state-of-the art is outlined in this paper

  2. Fission-gas bubble modeling for LMFBR accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1977-01-01

    The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

  3. Assessment of accident energetics in LMFBR core-disruptive accidents

    International Nuclear Information System (INIS)

    Fauske, H.K.

    1977-01-01

    An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

  4. Small leak shutdown, location, and behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Sandusky, D.W.

    1976-01-01

    The paper summarizes an experimental study of small leaks tested under LMFBR steam generator conditions. Defected tubes were exposed to flowing sodium and steam. The observed behavior of the defected tubes is reported along with test results of shutdown methods. Leak location methods were investigated. Methods were identified to open plugged defects for helium leak testing and detect plugged leaks by nondestructive testing

  5. Studies of LMFBR: method of analysis and some results

    International Nuclear Information System (INIS)

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

    1983-01-01

    Some results of recent studies of LMFBR characteristics are summarized. A two-dimensional model of the LMFBR is taken from a publication and used as the base model for the analysis. Axial structures are added to the base model and a three-dimensional (Δ - Z) calculation has been done. Two dimensional (Δ and RZ) calculations are compared with the three-dimensional and published results. The eigenvalue, flux and power distributions, breeding characteristics, control rod worth, sodium-void and Doppler reactivities are analysed. Calculations are done by CITATION using six-group cross sections collapsed regionwise by EXPANDA in one-dimensional geometries from the 70-group JFS library. Burnup calculations of a simplified thorium-cycle LMFBR have also been done in the RZ geometry. Principal results of the studies are: (1) the JFS library appears adequate for predicting overall characteristics of an LMFBR, (2) the sodium void reactivity is negative within - 25 cm from the outer boundary of the core, (3) the halflife of Pa-233 must be considered explicitly in burnup analyses, and (4) two-dimensional (RZ and Δ) calculations can be used iteratively to analyze three-dimensional reactor systems. (Author) [pt

  6. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

    Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

    1994-01-01

    In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

  7. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  8. Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

    International Nuclear Information System (INIS)

    Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

    1985-01-01

    Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

  9. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

    Anderson, R.; Hayden, O.

    2002-01-01

    Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

  10. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

    International Nuclear Information System (INIS)

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

    1977-01-01

    Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

  11. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

    The objectives of the Blanket Comparison and Selection Study (BCSS) can be stated as follows: (1) Define a small number (approx. 3) of blanket design concepts that should be the focus of the blanket R and D program. A design concept is defined by the selection of all materials (e.g., breeder, coolant, structure and multiplier) and other major characteristics that significantly influence the R and D requirements. (2) Identify and prioritize the critical issues for the leading blanket concepts. (3) Provide the technical input necessary to develop a blanket R and D program plan. Guidelines for prioritizing the R and D requirements include: (a) critical feasibility issues for the leading blanket concepts will receive the highest priority, and (b) for equally important feasibility issues, higher R and D priority will be given to those that require minimum cost and short time

  12. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

  13. Water tests for determining post voiding behavior in the LMFBR

    International Nuclear Information System (INIS)

    Hinkle, W.D.

    1976-06-01

    The most serious of the postulated accidents considered in the design of the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) is the Loss of Pipe Integrity (LOPI) accident. Analysis models used to calculate the consequences of this accident assume that once boiling is initiated film dryout occurs in the hot assembly as a result of rapid vapor bubble growth and consequent flow stoppage or reversal. However, this assumption has not been put to any real test. Once boiling is initiated in the hot assembly during an LMFBR LOPI accident, a substantial gravity pressure difference would exist between this assembly and other colder assemblies in the core. This condition would give rise to natural circulation flow boiling accompanied by pressure and flow oscillations. It is possible that such oscillations could prevent or delay dryout and provide substantial post-voiding heat removal. The tests described were conceived with the objective of obtaining basic information and data relating to this possibility

  14. Hydrogen jet recombination under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Wierman, R.W.

    1977-01-01

    Certain conditions may be postulated in LMFBR risk assessments for which the potential of hydrogen release to the reactor containment building needs to be evaluated. The inherent self-ignition characteristics of hydrogen jets entering the air atmosphere of the reactor containment building should be understood for such analyses. If hydrogen jets were to self-ignite (recombine) at the source where they enter the reactor containment building, then undesirable hydrogen accumulation would not occur. Therefore, experiments have been conducted investigating the phenomena associated with the recombination of hydrogen jets under conditions similar to those postulated for LMFBR studies. The data presented define the conditions required for self-ignition of the hydrogen jets

  15. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

    Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

    1986-01-01

    A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

  16. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

    Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Usher, J.L.

    1980-04-01

    Definitions are given of 10 generic blanket types and the specific blanket chosen to be analyzed in detail from each of the 10 types. Dimensions, compositions, energy depositions and breeding ratios (where applicable) are presented for each of the 10 designs. Ultimately, based largely on neutronics and thermal hyraulics results, breeding an nonbreeding blanket options are selected for further design analysis and integration with a suitable power conversion subsystem

  17. Cover-gas seals: 11-LMFBR seal-test program

    International Nuclear Information System (INIS)

    Steele, O.P. III; Horton, P.H.

    1977-01-01

    The objective of the Cover Gas Seal Material Development Program is to perform the engineering development required to provide reliable seals for LMFBR application. Specific objectives are to verify the performance of commercial solid cross-section and inflatable seals under reactor environments including radiation, to develop advanced materials and configurations capable of achieving significant improvement in radioactive gas containment and seal temperature capabilities, and to optimize seal geometry for maximum reliability and minimal gas permeation

  18. Axial migratin of cesium in LMFBR fuel pins

    International Nuclear Information System (INIS)

    Karnesky, R.A.; Bridges, A.E.; Jost, J.W.

    1981-11-01

    A correlated model for quantitatively predicting the behavior of cesium in LMFBR fuel pins has been developed. This correlation was shown to be in good agreement with experimental data. It has been used to predict the behavior of cesium in the FFTF driver fuel and as the result of this analysis it has been shown that the accumulation of cesium in the insulator pellets at the ends of the fuel column will not be life limiting

  19. Retention of gaseous fission products in reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Burch, W.D.; Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-05-01

    The report is devoted to status of the development programme at the Oak Ridge National Laboratory on methods for retaining iodine-131 and 129, Krypton-85, Tritium and Carbon-14 in reprocessing LMFBR fuels. The Iodox process, Fluorocarbon absorption process and Voloxidation process are described for retention of iodine, Krypton-85 and Tritium, respectively. Flowsheets for the different processes are given and results of experimental runs in small engineering-scale equipment are reported

  20. Future development LMFBR-steam generators SNR2

    International Nuclear Information System (INIS)

    Essebaggers, J.; Pors, J.G.

    1975-01-01

    The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

  1. Development of acidic processes for decontaminating LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    Hill, E F [Rockwell International, Atomics International Division, Canoga Park (United States); Colburn, R P; Lutton, J M; Maffei, H P [Hanford Engineering Development Laboratory, Richland (United States)

    1978-08-01

    The objective of the DOE decontamination program is to develop a well characterized chemical decontamination process for application to LMFBR primary system components that subsequently permits contact maintenance and allows requalification of the components for reuse in reactors. The paper describes the subtasks of deposit characterization, development of requalification and process acceptance criteria, development of process evaluation techniques and studies which led to a new acidic process for decontaminating 304 stainless steel hot leg components.

  2. LMFBR operational safety: the EBR-II experience

    International Nuclear Information System (INIS)

    Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

    1978-01-01

    The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

  3. Impact of LMFBR operating experience on PFBR design

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chetal, S.C.; Chellapandi, P.; Govindarajan, S.; Lee, S.M.; Kameswara Rao, A.S.L.; Prabhakar, R.; Raghupathy, S.; Sodhi, B.S.; Sundaramoorthy, T.R.; Vaidyanathan, G.

    2000-01-01

    PFBR is a 500 MWe, sodium cooled, pool type, fast breeder reactor currently under detailed design. It is essential to reduce the capital cost of PFBR in order to make it competitive with thermal reactors. Operating experience of LMFBRs provides a vital input towards simplification of the design, improving its reliability, enhancing safety and achieving overall cost reduction. This paper includes a summary of LMFBR operating experience and details the design features of PFBR as influenced by operating experience of LMFBRs. (author)

  4. Design approaches to achieve competitive LMFBR capital costs

    International Nuclear Information System (INIS)

    Arnold, W.H.; Ehrman, C.S.; Sharbaugh, J.E.; Young, W.H.

    1982-01-01

    Through analysis of the essential functional elements of an LMFBR, numerous ways were found to simplify system design, reduce the size of components and equipment, and eliminate some components and systems. The projected capital cost per net kW of this design is competitive with that of current PWRs. RandD programs and the construction and operation of CRBRP now are needed to prove out the features of this new design

  5. Thermal analysis methods for LMFBR wire wrapped bundles

    International Nuclear Information System (INIS)

    Todreas, N.E.

    1976-11-01

    A note is presented which was written to stimulate an awareness and discussion of the fundamental differences in the formulation of certain existing analysis codes for LMFBR wire wrap bundles. The contention of the note is that for those array types where data exists (one wire per pin, equal start angles), the ENERGY method results for coolant temperature under forced convection conditions provide benchmarks of reliability equal to the results of codes COBRA and TH1-3D

  6. Structural analysis for elevated temperature design of the LMFBR

    International Nuclear Information System (INIS)

    Griffin, D.S.

    1976-02-01

    In the structural design of LMFBR components for elevated temperature service it is necessary to take account of the time-dependent, creep behavior of materials. The accommodation of creep to assure design reliability has required (1) development of new design limits and criteria, (2) development of more detailed representations of material behavior, and (3) application of the most advanced analysis techniques. These developments are summarized and examples are given to illustrate the current state of technology in elevated temperature design

  7. Analytical work on local faults in LMFBR subassembly

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Miyaguchi, K.; Hirata, N.; Kasahara, F.

    1979-01-01

    Analytical codes have been developed for evaluating various severe but highly unlikely events of local faults in the LMFBR subassembly (S/A). These include: (1) local flow blockage, (2) two-phase thermohydraulics under fission gas release, and (3) inter-S/A failure propagation. A simple inter-S/A thermal failure propagation analysis code, FUMES, is described that allows an easy parametric study of propagation potential of fuel fog in a S/A. 7 refs

  8. A new approach to the design of LMFBR liners

    International Nuclear Information System (INIS)

    Polentz, L.M.

    1980-01-01

    An advance in the state-of-the-art of LMFBR liners which permits notable savings in construction costs without any sacrifice of safety is described. The application of the new design concept to the rework of the upper reactor vault liner of the FFTF is discussed. Factors which affect the application of the new design approach to other LMFBRs are delineated and discussed. (author)

  9. Effect of operating temperature on LMFBR core performance

    International Nuclear Information System (INIS)

    Noyes, R.C.; Bergeron, R.J.; di Lauro, G.F.; Kulwich, M.R.; Stuteville, D.W.

    1977-01-01

    The purpose of the study is to provide an engineering evaluation of high and low temperature LMFBR core designs. The study was conducted by C-E supported by HEDL expertise in the areas of materials behavior, fuel performance and fabrication/fuel cycle cost. The evaluation is based primarily on designs and analyses prepared by AI, GE and WARD during Phase I of the PLBR studies

  10. Shielding design method for LMFBR validation on the Phenix factor

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

    1983-05-01

    Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

  11. LMFBR technology. FFTF cover-gas leakage calculation

    International Nuclear Information System (INIS)

    Deboi, H.

    1974-01-01

    The FFTF LMFBR is intended to have a near zero release of radioactive gases during normal reactor operation with 1% failed fuel. This report presents calculations which provide an approximation of these cover gas leakages. Data from ongoing static and dynamic seal leak tests at AI are utilized. Leakage through both elastomeric and metallic seals in all sub-assemblies and penetrations comprising the reactor cover gas containment during reactor operation system are included

  12. Development of acidic processes for decontaminating LMFBR components

    International Nuclear Information System (INIS)

    Hill, E.F.; Colburn, R.P.; Lutton, J.M.; Maffei, H.P.

    1978-01-01

    The objective of the DOE decontamination program is to develop a well characterized chemical decontamination process for application to LMFBR primary system components that subsequently permits contact maintenance and allows requalification of the components for reuse in reactors. The paper describes the subtasks of deposit characterization, development of requalification and process acceptance criteria, development of process evaluation techniques and studies which led to a new acidic process for decontaminating 304 stainless steel hot leg components

  13. Biological behavior of mixed LMFBR-fuel-sodium aerosols

    International Nuclear Information System (INIS)

    Mahlum, D.D.; Hackett, P.L.; Hess, J.O.; Allen, M.D.

    1979-01-01

    Immediately after exposure of rats to mixed aerosols of sodium-LMFBR fuel, about 80 to 90% of the body burden of 239 Pu is in the gastrointestinal tract; 1.5 to 4% is in the lungs. With fuel-only aerosols, less of the body burden was in the GI tract and more in the lung and the head. Blood and urine values suggest an increased absorption of 239 Pu from sodium-fuel than from fuel-only aerosols

  14. Transport-diffusion comparisons for small core LMFBR disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1977-11-01

    A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for homogeneous cores and to the ZPPR-7 configurations for heterogeneous core. In all TUC cases diffusion theory was shown to be inadequate for the calculation of reactivity changes during core disassembly

  15. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  16. Research of the Mass Spectra of the Fission Products and Yields of (n, gamma) and (n, 2n) Reactions in a Model Subcritical Uranium Blanket of the Electronuclear System "Energy Plus Transmutation" on Proton Beam of the Dubna Synchrophasotron at 1.5 Ge

    CERN Document Server

    Chultem, D; Krivopustov, M I; Gerbish, S; Tumendemberel, B; Pavlyuk, A B; Zaveryukha, O S

    2002-01-01

    This paper is devoted to the research of the spatial distributions of the yields of (n, f), (n, gamma) and (n, 2n) reactions in a two-section model of the uranium blanket electronuclear installation constructed at the Laboratory of High Energies, JINR (Dubna) for experiments according to the program "Research of physical aspects of the electronuclear method of energy production and of radioactive waste transmutation in atomic power-engineering on beams of the synchrophasotron and nuclotron" - project "Energy plus Transmutation". The mass spectrum of the fission products and yields of above reactions in uranium activation detectors placed on the radii of the so-called detector plates is determined. The experimental results testify that the fission of nuclei in the uranium blanket is made by fast neutrons. This conclusion coincides with the result obtained with track integrators of uranium fission.

  17. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    Kleefeldt, K.

    1985-01-01

    The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m 2 . Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

  18. Market dynamics as a driver towards the evolution of research needs; the case of up-flow anaerobic sludge blanket seeding granules

    CSIR Research Space (South Africa)

    Musee, N

    2013-01-01

    Full Text Available Market dynamics offer positive (incentive) or negative (disincentive) feedback loops that shape the research needs for, or certain aspects of, a particular technology. Our case study results illustrate how market dynamics have influenced...

  19. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.; Khan, S.; Sahin, S.

    1983-01-01

    Recently, a project has started, with the aim to establish the neutronic performance and the basic design of an experimental fusionfission (hybrid) reactor facility, called AYMAN, in cylinderical geometry. The fusion reactor will have to be simulated by a (D,T) neutron generator. Fissile and fertile fuel will have to surround the neutron generator as a cylinderical blanket to simulate the boundary conditions of the hybrid blanket in a proper way. This geometry is consistent with Tandem Mirror Hybrid Blanket design and with most of the ICF blanket designs. A similar experimental installation will become operational around 1984 at the Swiss Federal Institute of Technology in Lausanne, Switzerland known under the project LOTUS. Due to the limited dimensions of the experimental cavity of the LOTUS-hybrid reactor, the LOTUS blankets have to be designed in plane geometry. Also, the bulky form of the Haefely neutron generator of the LOTUS facility obliges one to design a blanket in the plane geometry. This results in a vacuum left boundary conditions for the LOTUS blanket. The importance of a reflecting left boundary condition on the overall neutronic performance of a hybrid blanket has been analyzed in previous work in detail

  20. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-12-01

    This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

  1. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

  2. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

    A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

  3. Feasibility study of incore fission chamber application for neutron flux measurements on the NET blanket

    International Nuclear Information System (INIS)

    Bertalot, L.

    1987-01-01

    A feasibility study has been carried out on the use of in-core fission chambers as neutron diagnostic tools to perform neutron flux measurements on the blanket component of NET. The high neutron and gamma fluxes and the severe thermal-mechanical and magnetic conditions of the blanket structure have been taken into account in this analysis. Preliminary design criteria and specifications of an in-core detector are presented for NET application. A research and development programme is outlined which aims to obtain more information on the tecnological constraints arising from the severe conditions of the NET blanket

  4. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

    The NET machine has three horizontal ports reserved for testing tritium breeding blanket designs during the physics phase and possibly five during the technology phase. The design of the ports and test blankets are modular to accept a range of blanket options, provide radiation shielding and allow routine replacement. Radiation levels during replacement or maintenance require that all operations must be carried out remotely. The paper describes the problems overcome in providing a port design which includes attachment to the vacuum vessel with double vacuum seals, an integrated cooled first wall and support guides for the test blanket module. The method selected to remotely replace the test module whilst controlling the spread of contamination is also adressed. The paper concludes that the provisions of a test blanket facility based on the NET machine design is feasible. (orig.)

  5. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

    Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

  6. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

    This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

  7. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  8. Status of the LMFBR thermo- and fluid-dynamic activities at KFK

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hofmann, F.; Rehme, K.

    1979-01-01

    The aim of the thermo- and fluiddynamic analysis is to determine the spatial velocity and temperature distributions in LMFBR-core elements with high accuracy. Knowledge of these data is a necessary prerequisite for determining the mechanical behavior of fuel rods and of structural material. Three cases are distinguished: Nominal geometry and steady state conditions; non-nominal geometry and quasi-steady state conditions; nominal geometry and non-steady state conditions. The present situation for the design calculations of fuel elements is based mainly on undisturbed normal operation. Most of the thermo- and fluiddynamic activities performed under the Fast Breeder Programme at KFK are related to this case. The present status of theoretical and experimental research work briefly presented in this paper, can be subdivided into the following main topics: 1. Physical and mathematical modelling of single phase rod bundle thermo- and fluiddynamics, 2. Experimental investigations on heat transfer and fluid flow in rod bundles

  9. Peningkatan mutu blanket karet alam melalui proses predrying dan penyemprotan asap cair

    Directory of Open Access Journals (Sweden)

    Afrizal Vachlepi

    2017-06-01

    Full Text Available Most of Indonesian rubber products SIR 20 are made from the material of raw rubber obtained from smallholders. However, the quality of this material is not good enough. Thus, quality improvement has to be carried out by manufacturers. The liquid smoke used during the blanket hanging process can improve the quality of the rubber products SIR 20. This research aimed to determine and study the effects of liquid smoke spraying and blanket hanging duration on the drying factor, the dry rubber content, technical quality, vulcanization characteristics, and physical properties of vulcanized natural rubber. Treatments consisted of various hanging duration (6, 8, and 10 days, and without hanging and spraying (with and without spraying of liquid smoke. The results showed that the spraying of liquid smoke on natural rubber blankets could improve the technical quality of the natural rubber, especially the values of Po and PRI. The spraying of liquid smoke could reduce the blanket hanging duration to 6-8 days. The blankets sprayed with liquid smoke had the optimum cure time of around 15 minutes and 19 seconds and the scorch time of around 3 minutes and 22 seconds. These values indicated that the vulcanization characteristics of blankets which were sprayed with liquid smoke were generally better than those of blankets which were not sprayed with liquid smoke

  10. A development of user-friendly graphical interface for a blanket simulator

    International Nuclear Information System (INIS)

    Lee, Young-Seok; Yoon, Seok-Heun; Han, Jung-Hoon

    2010-01-01

    A web-based user-friendly graphical interface (GUI) system, named GUMBIS (Graphical User-friendly Monte-Carlo-Application Blanket-Design Interface System), was developed to cut down the efforts of the researchers and practitioners who study tokamak blanket designs with the Monte Carlo MCNP/MCNPX codes. GUMBIS was also aimed at supporting them to use the codes for their study without having through understanding on the complex menus and commands of the codes. Developed on the web-based environment, GUMBIS provides task sharing capability on a network. GUMBIS, applicable for both blanket design and neutronics analysis, could facilitate not only advanced blanket R and D but also the education and training of the researchers in the R and D.

  11. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

  12. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

    The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design

  13. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  14. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  15. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

    WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

    2002-01-01

    OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

  16. LMFBR models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1981-10-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given

  17. Evaluation of integrally finned cladding for LMFBR fuel pins

    International Nuclear Information System (INIS)

    Cantley, D.A.; Sutherland, W.H.

    1975-01-01

    An integral fin design effectively reduces the coolant temperature gradients within an LMFBR subassembly by redistributing coolant flow so as to reduce the maximum cladding temperature and increase the duct wall temperature. The reduced cladding temperatures are offset by strain concentrations resulting from the fin geometry, so there is little net effect on predicted fuel pin performance. The increased duct wall temperatures, however, significantly reduce the duct design lifetime so that the final conclusion is that the integral fin design is inferior to the standard wire wrap design. This result, however, is dependent upon the material correlations used. Advanced alloys with improved irradiation properties could alter this conclusion

  18. Fission and corrosion products behavior in primary circuits of LMFBR's

    International Nuclear Information System (INIS)

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  19. Confirmatory simulation of safety and operational transients in LMFBR systems

    International Nuclear Information System (INIS)

    Guppy, J.G.; Agrawal, A.K.

    1978-01-01

    Operational and safety transients that may originate anywhere in an LMFBR system must be adequately simulated to assist in safety evaluation and plant design efforts. This paper describes an advanced thermohydraulic transient code, the Super System Code (SSC), that may be used for confirmatory safety evaluations of plant wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions, and presents results obtained with SSC illustrating the degree of modelling detail present in the code as well as the computing efficiency. (author)

  20. LMFBR steam generator leak detection development in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Magee, P M; Gerrels, E E; Greene, D A [General Electric Company, Sunnyvale, CA (United States); McKee, J [Argonne National Laboratory, Argonne, IL (United States)

    1978-10-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H{sub 2} and O{sub 2}) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  1. Immersed acoustical transducers and their potential uses in LMFBR

    International Nuclear Information System (INIS)

    Argous, J.P.; Brunet, M.; Baron, J.; Lhuillier, C.; Segui, J.L.

    1980-04-01

    Six years satisfactory operation in PHENIX has proved the reliability and effectivness of under-sodium viewing (VISUS) and Acoustic Detection. This fact has been strong incentive to maintain, on the future LMFBR the visus as well as the Acoustic Detection functions. These two functions are performed on SUPER PHENIX, by two sets of distinct systems using the well-known solution. Taking into account of recent improvements in sodium immersible acoustic transducers technology, CEA decided to undertake the development of a multi-functions instrument. This paper gives an outline of this new concept, which should be able to reduce the cost and the complexity of core instrumentation

  2. CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

    1983-01-01

    The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

  3. LMFBR steam generator leak detection development in the United States

    International Nuclear Information System (INIS)

    Magee, P.M.; Gerrels, E.E.; Greene, D.A.; McKee, J.

    1978-01-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H 2 and O 2 ) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  4. Sodium mists behavior in cover gas space of an LMFBR

    International Nuclear Information System (INIS)

    Himeno, Y.; Takahashi, J.

    1978-03-01

    This paper present the sodium mist behavior in Argon cover gas space of an LMFBR experimentally using a test vessel of 1,400 mm in axial length, 305.5 mm in inner diameter and about 100 l in volume. Experiments are consisted with measurements of the mist concentration and the mist gravitational settling flux between the sodium pool temperature range of 290 0 to 520 0 C. The results are discussed under the monosize assumption of the particles, and the particle sizes and evaporation rate are derived. Transient and steady state mist concentration behavior were also investigated. (author)

  5. LMFBR fuel cycle studies progress report, August 1972, No. 42

    International Nuclear Information System (INIS)

    Unger, W.E.; Blanco, R.E.; Crouse, D.J.; Irvine, A.R.; Watson, C.D.

    1972-10-01

    This report continues a series outlining progress in the development of methods for reprocessing of LMFBR fuels. Development work is reported on problems of irradiated fuel transport to the processing facility, the dissolution of the fuel and the chemical recovery of PuO 2 --UO 2 values, the containment of volatile fission products, product purification, conversion of fuel processing plant product nitrate solutions to solids suitable for shipping and for subsequent fuel fabrication. Pertinent experimental results are presented for the information of those immediately concerned with the field. Detailed description of experimental work and data are included in the topical reports and in the Chemical Technology Division Annual Reports

  6. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  7. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kuroda, Toshimasa; Enoeda, Mikio; Kikuchi, Shigeto [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-11-01

    The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

  8. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-10-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

  9. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

    1998-01-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

  10. Beryllium R&D for blanket application

    Science.gov (United States)

    Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

    1998-10-01

    The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

  11. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1978-01-01

    The thermal and mechanical aspects of hybrid reactor blanket design considerations are discussed. This paper is intended as a companion to that of J. D. Lee of Lawrence Livermore Laboratory on the nuclear aspects of hybrid reactor blanket design. The major design characteristics of hybrid reactor blankets are discussed with emphasis on the areas of difference between hybrid reactors and standard fusion or fission reactors. Specific examples are used to illustrate the design tradeoffs and choices that must be made in hybrid reactor design. These examples are drawn from the work on the Mirror Hybrid Reactor

  12. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.E.; Cheng, E.T.

    1985-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

  13. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.L.; Cheng, E.T.

    1986-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

  14. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Young, J.R.

    1975-01-01

    Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

  15. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

  16. Blanket design for imploding liner systems

    International Nuclear Information System (INIS)

    Schaffer, M. J.

    1980-01-01

    The blanket design comprises hot, molten, rotating liquid vortex systems suitable for rapidly compressing confined plasmas, in which stratified immiscible liquid layers having successively greater mass densities outwardly of the axis of rotation are provided

  17. APT target-blanket fabrication development

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.L.

    1997-06-13

    Concepts for producing tritium in an accelerator were translated into hardware for engineering studies of tritium generation, heat transfer, and effects of proton-neutron flux on materials. Small-scale target- blanket assemblies were fabricated and material samples prepared for these performance tests. Blanket assemblies utilize composite aluminum-lead modules, the two primary materials of the blanket. Several approaches are being investigated to produce large-scale assemblies, developing fabrication and assembly methods for their commercial manufacture. Small-scale target-blanket assemblies, designed and fabricated at the Savannah River Site, were place in Los Alamos Neutron Science Center (LANSCE) for irradiation. They were subjected to neutron flux for nine months during 1996-97. Coincident with this test was the development of production methods for large- scale modules. Increasing module size presented challenges that required new methods to be developed for fabrication and assembly. After development, these methods were demonstrated by fabricating and assembling two production-scale modules.

  18. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

    The design features of the hybrid reactor blanket and main parameters are presented. The stress analysis is performed by using computer codes SAP5p and SAP6 with the three kinds of blanket module loadings, i.e, the pressure of coolant, the blanket weight and the thermal loading. Numerical calculation results indicate that the stresses of the blanket are smaller than the allowable ones of the material, the blanket design is therefore reasonable

  19. Analytical approach for confirming the achievement of LMFBR reliability goals

    International Nuclear Information System (INIS)

    Ingram, G.E.; Elerath, J.G.; Wood, A.P.

    1981-01-01

    The approach, recommended by GE-ARSD, for confirming the achievement of LMFBR reliability goals relies upon a comprehensive understanding of the physical and operational characteristics of the system and the environments to which the system will be subjected during its operational life. This kind of understanding is required for an approach based on system hardware testing or analyses, as recommended in this report. However, for a system as complex and expensive as the LMFBR, an approach which relies primarily on system hardware testing would be prohibitive both in cost and time to obtain the required system reliability test information. By using an analytical approach, results of tests (reliability and functional) at a low level within the specific system of interest, as well as results from other similar systems can be used to form the data base for confirming the achievement of the system reliability goals. This data, along with information relating to the design characteristics and operating environments of the specific system, will be used in the assessment of the system's reliability

  20. Feasibility study on large pool-type LMFBR

    International Nuclear Information System (INIS)

    1984-01-01

    A feasibility study has been conducted from 1981 FY to 1983 FY, in order to evaluate the feasibility of a large pool-type LMFBR under the Japanese seismic design condition and safety design condition, etc. This study was aimed to establish an original reactor structure concept which meets those design conditions especially required in Japan. In the first year, preceding design concepts had been reviewed and several concepts were originated to be suitable to Japan. For typical two of them being selected by preliminary analysis, test programs were planned. In the second year, more than twenty tests with basic models had been conducted under severe conditions, concurrently analytical approaches were promoted. In the last year, larger model tests were conducted and analytical methods have been verified concerning hydrodynamic effects on structure vibration, thermo-hydraulic behaviours in reactor plena and so on. Finally the reactor structure concepts for a large pool-type LMFBR have been acknowledged to be feasible in Japan. (author)

  1. Proposal for computer investigation of LMFBR core meltdown accidents

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

    1974-01-01

    The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

  2. Confirmatory simulation of safety and operational transients in LMFBR systems

    International Nuclear Information System (INIS)

    Guppy, J.G.; Agrawal, A.K.

    1978-01-01

    Operational and safety transients (anticipated, unlikely, or extremely unlikely) that may originate anywhere in a liquid-metal fast breeder reactor (LMFBR) system must be adequately simulated to assist in safety evaluation and plant design efforts. An advanced thermohydraulic transient code, the Super System Code (SSC), is described that may be used for confirmatory safety evaluations of plant-wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions. Results obtained with SSC illustrating the degree of modeling detail present in the code as well as the computing efficiency are presented. A version of the SSC code, SSC-L, applicable to any loop-type LMFBR design, has been developed at Brookhaven. The scope of SSC-L is to enable the simulation of all plant-wide transients covered by Plant Protection System (PPS) action, including sodium pipe rupture and coastdown to natural circulation conditions. The computations are stopped when loss of core integrity (i.e., clad melting temperature exceeded) is indicated

  3. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    Broc, Daniel

    2001-01-01

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  4. LMFBR core flowering response to an impulse load

    International Nuclear Information System (INIS)

    Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

    1993-01-01

    Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

  5. LMFBR subassembly response to simulated local pressure loadings

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1976-01-01

    The structural response of liquid metal fast breeder reactor (LMFBR) subassemblies to local accidental events is of interest in assessing the safety of such systems. Problems to be resolved include failure propagation modes from pin to pin and from subassembly to subassembly. Factors which must be considered include: (a) the geometry of the structure, (b) uncertainty of the pressure-energy source, (c) uncertainty of materials properties under reactor operating conditions, and (d) the difficulty in performing in-pile or out-of-pile experiments which would simulate the above conditions. The main effort in evaluating the subassembly response has been centered around the development of appropriate analyses based on the finite element technique. Analysis has been extended to include not only the subassembly duct structure itself, but also the fluid environment, both within subassemblies and between them. These models and codes have been devised to cover a wide range of accident loading conditions, and can treat various materials as their properties become known. The effort described here is centered mainly around an experimental effort aimed at verfying, modifying or extending the models used in treating subassembly damage propagation. To verify the finite element codes under development, a series of out-of-pile room temperature experiments has been performed on LMFBR-type subassembly ducts under various loading conditions. (Auth.)

  6. Transient analysis of LMFBR reinforced/prestressed concrete containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Belytschko, T.B.; Bazant, Z.P.

    1979-01-01

    The use of prestressed concrete reactor vessels (PCRVs) for LMFBR containment creates a need for analytical methods for treating the transient response of such structures, for LMFBR containments must be capable of sustaining the dynamic effects which arise in a hypothetical core disruptive accident (HCDA). These analyses require several unique features: a model of concrete which includes tensile cracking, a methodology for representing the prestressing tendons and for simulating the prestressing operation, and an efficient computational tool for treating the transient response. Furthermore, for the sake of convenience, all of these features should be available in a single computer code. For the purpose of treating the transient response, a finite element program with explicit time integration was chosen. The use of explicit time integration has the advantage that it can easily treat the complicated constitutive model which arises from the considerations of concrete cracking and it can handle the slip between reinforcing tendons and the concrete through the use of the well known sliding interface options. However, explicit time integration programs are usually not well suited to the simulation of static processes such as prestressing. Nevertheless, explicit time integration programs can handle static processes through the introduction of damping by what is known as a dynamic relaxation procedure. For this reason, the dynamic relaxation procedure was refined through the introduction of lumped mass, viscous damping. This provision made the prestressing operation of the concrete structures by means of the explicit formulation rather convenient. (orig.)

  7. Rapid thermal cycling of new technology solar array blanket coupons

    Science.gov (United States)

    Scheiman, David A.; Smith, Bryan K.; Kurland, Richard M.; Mesch, Hans G.

    1990-01-01

    NASA Lewis Research Center is conducting thermal cycle testing of a new solar array blanket technologies. These technologies include test coupons for Space Station Freedom (SSF) and the advanced photovoltaic solar array (APSA). The objective of this testing is to demonstrate the durability or operational lifetime of the solar array interconnect design and blanket technology within a low earth orbit (LEO) or geosynchronous earth orbit (GEO) thermal cycling environment. Both the SSF and the APSA array survived all rapid thermal cycling with little or no degradation in peak performance. This testing includes an equivalent of 15 years in LEO for SSF test coupons and 30 years of GEO plus ten years of LEO for the APSA test coupon. It is concluded that both the parallel gap welding of the SSF interconnects and the soldering of the APSA interconnects are adequately designed to handle the thermal stresses of space environment temperature extremes.

  8. The fusion blanket program at Chalk River

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-03-01

    Work on the Fusion Blanket Program commenced at Chalk River in 1984 June. Co-funded by Canadian Fusion Fuels Technology Project and Atomic Energy of Canada Limited, the Program utilizes Chalk River expertise in instrumented irradiation testing, ceramics, tritium technology, materials testing and compound chemistry. This paper gives highlights of studies to date on lithium-based ceramics, leading contenders for the fusion blanket

  9. Failed fuel detection and location of LMFBR

    International Nuclear Information System (INIS)

    Mimoto, Yasuhide; Hukuda, Tooru; Nakamoto, Koichiro

    1974-01-01

    This is a summary report on Failed Fuel Detection and Location Methods of liquid metal cooled fast breeder reactors, and describes an outline of related research and development conducted by PNC. (auth.)

  10. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Malang, S.; Nishio, S.; Raffray, R.; Sagara, A.

    2002-01-01

    First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

  11. Thermal comfort and safety of cotton blankets warmed at 130°F and 200°F.

    Science.gov (United States)

    Kelly, Patricia A; Cooper, Susan K; Krogh, Mary L; Morse, Elizabeth C; Crandall, Craig G; Winslow, Elizabeth H; Balluck, Julie P

    2013-12-01

    In 2009, the ECRI Institute recommended warming cotton blankets in cabinets set at 130°F or less. However, there is limited research to support the use of this cabinet temperature. To measure skin temperatures and thermal comfort in healthy volunteers before and after application of blankets warmed in cabinets set at 130 and 200°F, respectively, and to determine the time-dependent cooling of cotton blankets after removal from warming cabinets set at the two temperatures. Prospective, comparative, descriptive. Participants (n = 20) received one or two blankets warmed in 130 or 200°F cabinets. First, skin temperatures were measured, and thermal comfort reports were obtained at fixed timed intervals. Second, blanket temperatures (n = 10) were measured at fixed intervals after removal from the cabinets. No skin temperatures approached levels reported in the literature that cause epidermal damage. Thermal comfort reports supported using blankets from the 200°F cabinet, and blankets lost heat quickly over time. We recommend warming cotton blankets in cabinets set at 200°F or less to improve thermal comfort without compromising patient safety. Copyright © 2013 American Society of PeriAnesthesia Nurses. Published by Elsevier Inc. All rights reserved.

  12. Liquid lithium blanket processing studies

    International Nuclear Information System (INIS)

    Talbot, J.B.; Clinton, S.D.

    1979-01-01

    The sorption of tritium on yttrium from flowing molten lithium and the subsequent release of tritium from yttrium for regeneration of the metal sorbent were investigated to evaluate the feasibility of such a tritium-recovery process for a fusion reactor blanket of liquid lithium. In initial experiments with the forced convection loop, yttrium samples were contacted with lithium at 300 0 C. A mass transfer coefficient of 2.5 x 10 - cm/sec, which is more than an order of magnitude less than the value measured in earlier static experiments, was determined for the flowing lithium system. Rates of tritium release from yttrium samples were measured to evaluate possible thermal regeneration of the sorbent. Values for diffusion coefficients at 505, 800, and 900 0 C were estimated to be 1.1 x 10 -13 , 4.9 x 10 -12 , and 9.3 x 10 -10 cm 2 /sec, respectively. Tritium release from yttrium was investigated at higher temperatures and with hydrogen added to the argon sweep gas to provide a reducing atmosphere

  13. New approach to the design of core support structures for large LMFBR plants

    International Nuclear Information System (INIS)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions

  14. State of the art review of degradation processes in LMFBR materials. Volume II. Corrosion behavior

    International Nuclear Information System (INIS)

    Dillon, R.D.

    1975-01-01

    Degradation of materials exposed to Na in LMFBR service is reviewed. The degradation processes are discussed in sections on corrosion and mass transfer, erosion, wear and self welding, sodium--water reactions, and external corrosion. (JRD)

  15. Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

    International Nuclear Information System (INIS)

    Moreira, M.L.

    1985-01-01

    A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

  16. Seismic response and damping tests of small bore LMFBR piping and supports

    International Nuclear Information System (INIS)

    Barta, D.A.; Anderson, M.J.; Severud, L.K.; Lindquist, M.R.

    1984-01-01

    Seismic testing and analysis of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described. Measured responses to simulated seismic excitations are compared with analytical predictions based on NRC Regulatory Guide 1.61 and measured system damping values. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps

  17. LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Dhir, V.K.; Frank, M.; Kastenberg, W.E.; McKone, T.E.

    1979-03-01

    Three aspects of LMFBR safety are discussed. The first concerns the potential reactivity effects of whole core fuel motion prior to pin failure in low ramp rate transient overpower accidents. The second concerns the effects of flow blockages following pin failure on the coolability of a core following an unprotected overpower transient. The third aspect concerns the safety related implications of using thorium based fuels in LMFBR's

  18. Maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

    Verriere, P.; Alanche, J.; Minguet, J.L.

    1984-06-01

    After some general remarks on the French fast neutron system, this paper presents the state of the program for the construction of fast reactor in France. Then, the general design of Super Phenix 1 steam generator components is outlined and, the in-service monitoring systems and protective devices with which they are equiped are briefly described. The methods used, in the event of leakage, for leak location, steam generator inspection, steam generator repair and putting the affected loop back into service, are discussed. There are two main lines of research, relating respectively to the means of water leak detection in sodium and the inspection arrangements that will be used either periodically, or following a sodium-water reaction. Finally, after a brief description of the steam generator, this paper describes the four incidents (leaks) that occurred on the Phenix steam generator in the course of 1982 and 1983, and the subsequent repair operations

  19. Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

    International Nuclear Information System (INIS)

    Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

    1982-01-01

    One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

  20. Emergency air cleaning system development for LMFBR containments

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.; Muhlestein, L.D.

    1975-01-01

    Criteria for evaluating the various types of Emergency Air Cleaning Systems which may be used in LMFBR plants have been established for both single containment and containment-confinement arrangements. These two plant arrangements have quite different air cleaning requirements for postulated design base accident conditions. Work is currently in progress to select from a list of candidate air cleaning systems those which best meet the criteria requirements. By means of a weighted rating system, areas of strength or weakness can be found and the conceptual system design then optimized. The final system arrangements will be ranked and several of the most promising systems selected for large-scale tests in the former CSE vessel at Hanford. 8 references. (U.S.)

  1. Development of a simple estimation tool for LMFBR construction cost

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Kinoshita, Izumi

    1999-01-01

    A simple tool for estimating the construction costs of liquid-metal-cooled fast breeder reactors (LMFBRs), 'Simple Cost' was developed in this study. Simple Cost is based on a new estimation formula that can reduce the amount of design data required to estimate construction costs. Consequently, Simple cost can be used to estimate the construction costs of innovative LMFBR concepts for which detailed design has not been carried out. The results of test calculation show that Simple Cost provides cost estimations equivalent to those obtained with conventional methods within the range of plant power from 325 to 1500 MWe. Sensitivity analyses for typical design parameters were conducted using Simple Cost. The effects of four major parameters - reactor vessel diameter, core outlet temperature, sodium handling area and number of secondary loops - on the construction costs of LMFBRs were evaluated quantitatively. The results show that the reduction of sodium handling area is particularly effective in reducing construction costs. (author)

  2. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. A technique for computing bowing reactivity feedback in LMFBR's

    International Nuclear Information System (INIS)

    Finck, P.J.

    1987-01-01

    During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

  4. CAT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Betten, P.R.; Tow, D.M.

    1984-04-01

    A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

  5. Validation of turbulence models for LMFBR outlet plenum flows

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-01-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds number (Re) values of 33000 and 70000 in a 1/15 - scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different two-equation turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet flow field, importantly also upon the degree of inlet turbulence, and also upon the turbulent momentum exchange model used in the calculations. In the FFTF geometry, the TEACH-T predictions agree well with the experiments. 7 refs

  6. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800 0 C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800 0 C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800 0 C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (Δanti G 02 congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800 0 C

  7. Fatigue of LMFBR piping due to flow stratification

    Energy Technology Data Exchange (ETDEWEB)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface.

  8. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. (Auth.)

  9. Simple LMFBR axial-flow friction-factor correlation

    International Nuclear Information System (INIS)

    Chan, Y.N.; Todreas, N.E.

    1981-09-01

    Complicated LMFBR axial lead-length averaged friction factor correlations are reduced to an easy, ready-to-use function of bundle Reyonlds number for wire-wrapped bundles. The function together with the power curves to calculate the associated constants are incorporated in a computer pre-processor, EZFRIC. The constants required for the calculation of the subchannels and bundle friction factors are derived and correlated into power curves of geometrical parameters. A computer program, FRIC, which can alternatively be used to accurately calculate these constants is also included. The accuracte values of the constants and the corresponding values predicted by the power curves and percentage error of prediction are tabulated for a wide variety of geometries of interest

  10. Fatigue of LMFBR piping due to flow stratification

    International Nuclear Information System (INIS)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface

  11. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

    The ITER Test Blanket Working Group (TBWG) has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out until July 1998, which formed part of the ITER Final Design Report (FDR), completed in 1998. Whilst the machine parameters for ITER-FEAT have been significantly revised compared to the FDR, testing of breeding blanket modules remains a main objective of the test programme and the development of a reactor-relevant breeding blanket to ensure tritium fuel self-sufficiency is recognized a key issue for fusion. Design work and R and D on breeding blanket concepts, including co-operation with the other Contacting Parties of the ITER-EDA for testing these concepts in ITER, are included in the work plans of the Parties

  12. Creep strain accumulation in a typical LMFBR piperun

    International Nuclear Information System (INIS)

    Johnstone, T.L.

    1975-01-01

    The analysis described allows the strain concentrations in typical LMFBR two anchor point uniplanar piperuns to be calculated. Account is taken of the effect of pipe elbows in attracting creep strain to themselves as well as possible movements of the thrust line due to strain redistribution. The influence of the initial load conditions is also examined. The stress relaxation analysis is facilitated by making the assumption that a cross-sectional stress distribution determined by the asymptotic fully developed state of creep exists at all times. Use is then made of Hoff(s) analogy between materials with a creep law of the Norton type and those with a corresponding non-linear elastic stress strain law, to determine complementary strain energy rates for straight pipes and bends. Ovalisation of the latter produces an increased strain energy rate which can be simply calculated by comparison with an equal length of straight pipe through employing a creep flexibility factor due to Spence. Deflection rates at any location in the pipework can then be evaluated in terms of the thermal restraint forces at that location by an application of Castigliano's principle. In particular for an anchor point the deflection rates are identically zero and this leads to the generation of 3 simultaneous differential equations determining the relaxation of the anchor reactions. Indicative results are presented for the continuous relaxation at 570 deg C of the thermally induced stress in a planar approximation to a typical LMFBR pipe run chosen to have peak elbow stresses close to the code maximum. The results indicate a ratio, after 10 5 hours, of 3 for creep strain concentration relative to initial peak strain (calculated on the assumption of fully elastic behavior) in the most severely affected elbow, when either austenitic 316 or 321 creep properties are employed

  13. Study on the phenomena of natural circulation in LMFBR

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Koga, Tomonari

    1993-01-01

    Decay heat removal with natural circulation is to be introduced to the LMFBR operation under loss of the electric power supply. The natural circulation is highly reliable, but the phenomenon is essentially unstable and subtle, which makes fine prediction difficult. The difficulties of experimental prediction are explained by facts that the phenomena are ruled by the delicate balance between the buoyancy force and the low pressure loss and are influenced by the various parameters such as local geometry, heat capacity and so on. Therefore the similarity rule for the natural circulation has not been fully understood. This study has been conducted to establish the simulation method for the natural circulation phenomena and the detailed phenomena have been reviewed. For the natural circulation in an LMFBR plant, there are no readily available reference velocity and temperature. These values are related only with the heating and cooling rate, the characteristic length and physical properties of the testing fluid. Basic equations were transformed by these values, and dimensionless equations were derived and then two dimensionless numbers, the Gr' number and the Bo' number, were identified. In order to examine the similarity rule for natural circulation we performed experiments using the different scale water models, a 1/20th and a 1/6th model. The temperatures and velocities at typical points were measured in the transient condition with various heating rate as a parameter. Measured temperatures and velocities were transformed to dimensionless forms for comparison and the effects of the Bo' number and the Gr' number were examined. As a result, it was clarified that the effect of the Gr' number is negligibly small but the effect of Bo' number still remained in our experimental range. The Bo' number of an actual plant is within the range of this experiment. Accordingly similitude of the Bo' number becomes important in an experiment to simulate an actual plant. (author)

  14. Argon cover gas purity control on LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Maeda, Hiroshi; Kobayashi, Takayoshi [PNC (Japan); Ishiyama, Satoshi [Toshiba (Japan); Motonaga, Tetsuji [Hitachi (Japan)

    1987-07-01

    Various control methods on chemical impurities and radioactive materials (fission products) in the primary argon gas of LMFBRs' have been studied based on experiences in Joyo and results of research and development. These results are reflected on MONJU design. On-line gas chromatographs are installed both in the Primary and in the Secondary Argon Gas Systems in JOYO. Also, chemical analysis has been done by batch sampling in JOYO. Though the rise of impurity concentration had been measured after periodical fuel exchange operation, impurity concentration has been controlled sufficiently under target control limits. In MONJU detailed design, the Rare Gas Removal and Recovery System which consisted of cryogenic distillation equipment had been eliminated and the capacity of Charcoal Beds in the Primary Argon Gas System has been improved to keep the concentration of radioactive materials sufficient low levels. The necessity to control the impurities in fresh argon gas which is supplied to the Primary Argon Gas System is now considered to keep the concentration of Kr and Xe isotopes in specified level, because their isotopes may make background rise for the Tagging Gas Failed Fuel Detection and Location System. Based on various investigations performed on sodium vapor trapping to obtain its detailed characteristics, design specifications and operating conditions of MONJU's Vapor Traps have been decided. To keep the level of radioactivity in gaseous effluents to the environment as low as reasonably achievable, the following means are now adopted in MONJU: the Primary Argon Gas System is composed of a closed recirculating path, but the exhaust gas discharged has different path after the Charcoal Beds; fresh argon gas is blown down to prevent Primary Argon Gas from releasing to the circumference during opening of the primary argon gas boundary, such as fuel exchange operations. (author)

  15. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

    The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

  16. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1994-11-01

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

  17. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Palmer, B.J.F.

    1987-11-01

    There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

  18. The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse.

    1988-03-01

    We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

  19. Blanket/first wall challenges and required R&D on the pathway to DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-11-15

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  20. Blanket/first wall challenges and required R&D on the pathway to DEMO

    International Nuclear Information System (INIS)

    Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-01-01

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  1. Epoxy blanket protects milled part during explosive forming

    Science.gov (United States)

    1966-01-01

    Epoxy blanket protects chemically milled or machined sections of large, complex structural parts during explosive forming. The blanket uniformly covers all exposed surfaces and fills any voids to support and protect the entire part.

  2. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1981-01-01

    The Tandem Mirror Reactor, with its cylindrical central cell, has led to numerous blanket designs taking advantage of the simple geometry. Also many new applications for fusion neutrons are now being considered. To the pure fusion electricity producers and hybrids producing fissile fuel, we are adding studies of synthetic fuel producers and fission-suppressed hybrids. The three blanket concepts presented are new ideas and should be considered illustrative of the breadth of Livermore's application studies. They are not meant to imply fully analyzed designs

  3. Fusion blanket high-temperature heat transfer

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-01-01

    Deep penetration of 14 MeV neutrons makes two-temperature region blankets feasible. A relatively low-temperature (approx. 300 0 C) metallic structure is the vacuum/coolant pressure boundary, while the interior of the blanket, which is a simple packed bed of nonstructural material, operates at very high temperatures (>1000 0 C). The water-cooled shell structure is thermally insulated from the steam-cooled interior. High-temperature steam can dramatically increase the efficiency of electric power generation, as well as produce hydrogen and oxygen-based synthetic fuels at high-efficiency

  4. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

    Tritium release from the candidate ceramic materials, Li 2 O, LiA10 2 , Li 2 SiO 3 , Li 4 SiO 4 and Li 2 ZrO 3 , is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

  5. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  6. Processing and waste disposal needs for fusion breeder blankets system

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1988-01-01

    We evaluated the waste disposal and recycling requirements for two types of fusion breeder blanket (solid and liquid). The goal was to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under U.S. Nuclear Regulatory Commission regulations. Described in this paper are the radionuclides expected in fusion blanket materials, plans for reprocessing and disposal of blanket components, and estimates for the operating costs involved in waste disposal. (orig.)

  7. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

    This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

  8. Proceedings of the eleventh international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2004-07-01

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li 2 TiO 3 and so on, fabrication technology developments and characterization of the Li 2 TiO 3 and Li 4 SiO 4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li 2 TiO 3 and Li 4 SiO 4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  9. Long-range research plan: FY 1984-FY 1988

    International Nuclear Information System (INIS)

    1982-08-01

    Information is presented concerning planned research activities related to LOCA and transients; LOFT; accident evaluation and mitigation; LMFBR and HTGR type reactors; facility operations and safeguards; waste management; siting and environment; and system and reliability analysis

  10. A review of ANL base technology studies in support of the U.S. LMFBR vibration program

    International Nuclear Information System (INIS)

    Wambsganss, M.W.; Chen, S.S.; Mulcahy, T.M.; Shin, Y.S.

    1977-01-01

    Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

  11. A review of ANL base technology studies in support of the U.S. LMFBR vibration program

    Energy Technology Data Exchange (ETDEWEB)

    Wambsganss, M W; Chen, S S [Components Technology Division, Argonne National Laboratory, Argonne, IL (United States); Mulcahy, T M; Shin, Y S

    1977-12-01

    Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

  12. Accelerator-driven molten-salt blankets: Physics issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-01-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt, accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external, moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

  13. Accelerator-driven molten-salt blankets: Physics issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-01-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

  14. INTOR first wall/blanket/shield activity

    International Nuclear Information System (INIS)

    Gohar, Y.; Billone, M.C.; Cha, Y.S.; Finn, P.A.; Hassanein, A.M.; Liu, Y.Y.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.

    1986-01-01

    The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory

  15. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

    The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

  16. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  17. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Anzidei, L.

    1994-01-01

    There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development. (authors). 8 figs., 1 tab

  18. A blanket design, apparatus, and fabrication techniques for the mass production of multilayer insulation blankets for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.; Schoo, C.J.

    1989-09-01

    The multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) consists of full cryostat length assemblies of aluminized polyester film fabricated in the form of blankets and installed as blankets to the 4.5K cold mass and the 20K and 80K thermal radiation shields. Approximately 40,000 MLI blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket is nearly 17 meters long and 1.8 meters wide. This paper reports the blanket design, an apparatus, and the fabrication method used to mass produce pre-fabricated MLI blankets. Incorporated in the blanket design are techniques which automate quality control during installation of the MLI blankets in the SSC cryostat. The apparatus and blanket fabrication method insure consistency in the mass produced blankets by providing positive control of the dimensional parameters which contribute to the thermal performance of the MLI blanket. By virtue of the fabrication process, the MLI blankets have inherent features of dimensional stability three-dimensional uniformity, controlled layer density, layer-to-layer registration, interlayer cleanliness, and interlayer material to accommodate thermal contraction differences. 11 refs., 6 figs., 1 tab

  19. R and D status on Water Cooled Ceramic Breeder Blanket Technology

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji

    2014-10-15

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li{sub 2}TiO{sub 3} pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

  20. Interactions of D-T neutrons in graphite and lithium blankets of fusion reactors

    International Nuclear Information System (INIS)

    Ofek, R.

    1986-05-01

    The present study deals with integral experiment and calculation of neutron energy spectra in bulks of graphite which is used as a reflector in blankets of fusion reactors, and lithium, the material of the blanket on which lithium is bred due to neutron interactions. The collimated beam configuration enables - due to the almost monoenergeticity and unidirectionality of the neutrons impinging on the target - to identify fine details in the measured spectra, and also facilitates the absolute normalization of the spectra. The measured and calculated spectra are generally in a good agreement and in a very good agreement at mesh points close to the system axis. A few conclusions may be drawn: a) the collimated beam source configuration is a sensitive tool for measuring neutron energy spectra with a high resolution, b) the method of unfolding proton-recoil spectra measured with a NE-213 scintillator should be improved, c) MCNP and DOT 4.2 may be used as complementary codes for neutron transport calculations of fusion blankets and deep-penetration problems, d) the updating of the cross-section libraries and checking by integral experiments is highly important for the design of fusion blankets. The present study may be regarded as an important course in the research and development of tools for the design of fusion blankets

  1. A fundamental study on sodium-water reaction in the double pool LMFBR, (3)

    International Nuclear Information System (INIS)

    Uotani, Masaki; Kumagai, Hiromichi; Nishi, Yoshihisa; Yoshida, Kazuo

    1989-01-01

    The double pool LMFBR is an innovative reactor that Central Research Institute of Electric Power Industry proposed for the purpose of reducing the construction cost of FBRs, and it is characterized by immersing steam generators in the annular plenum formed between the primary vessel and the outer secondary vessel. Therefore, it is expected that the pressure behavior at the time of sodium-water reaction due to the breaking of heating tubes is largely different from the case of steam generators of conventional FBRs. In order to ensure the soundness of the primary vessel that containes the reactor core, it is necessary to sufficiently grasp the pressure behavior in the plenum, and this basic experiment and analysis are related to the pressure behavior due to piston motion that arises in the initial period of quasi-steady pressure. About 1/10 scale annular plenum was used, and the generation of reaction product gas was simulated by the release of nitrogen. When gas was released in the plenum, the highest pressure rise occurred in the initial period of release, and thereafter, periodic variation arose. The pressure waveform and the value of pressure rise as the results of the model analysis agreed well with the measured results. (K.I.)

  2. Review on Japanese activities in the field of maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

    Tsuchiya, T.; Fukuda, T.; Sato, M.; Okabayashi, K.; Takahashi, T.

    2002-01-01

    Summary of Japanese activities on maintenance and repair of LMFBR steam generators (SG) is described in this paper. The concept (adoption of helical coil tube etc.) of MONJU SG was established in conceptional design started from 1968, and research and development (R and D) program was prepared. Parallel with basic studies such as material, welding, sodium water reaction and etc., overall verification tests using mock up SGs were conducted. As the first step, 1 Mw SG with two active helical tubes (and eight dummy tubes) was fabricated and operated, and many maintenance and repair experiences were accumulated through two small water leak troubles. Two 50Mw SGs, 1/5 scale of MONJU SG, were constructed and operated for long time. Post test examinations were carried out for No.1 50 Mw SG and feasibility of this type of SG was confirmed. In regard to maintenance and repair techniques, explosive and welding method for tube plugging and UT and ECT techniques for inspection of tube integrity are under development. Overall verification test for on-site and in-factory maintenance and repair techniques was conducted using No.2 50Mw SG evaporator and applicability of those techniques to real plant was evaluated. Many experiences were accumulated for removal and cleaning of sodium water reaction products after sodium water reaction in the cooling system and pressure relief system, using the Large Sodium Water Reaction Test Facility (SWAT-1 and 3). (author)

  3. Development of computer code models for analysis of subassembly voiding in the LMFBR

    International Nuclear Information System (INIS)

    Hinkle, W.

    1979-12-01

    The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered

  4. Studies of flow stratification in the hot plenum of an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Jones, P; Hickmott, S [Central Electricity Generating Board, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire (United Kingdom)

    1983-07-01

    The paper reviews work at Berkeley Nuclear Laboratories on the extent and effects of buoyancy in the hot plenum of an LMFBR. It summarizes the experimental, theoretical and numerical work has has been conducted to aid the understanding of the complex transient flows which occur following a reactor trip. The experimental work has been conducted in small-scale idealised geometries which isolate the essential features of the reactor flows and is not intended to provide detailed design data. An integral theory has been devised to describe the thermal hydraulics of negatively-buoyant jets. The predictions are shown to be in good agreement with the experimental results and emphasize the need to correctly represent the inlet velocity and temperature profiles. Some preliminary calculations with a transient, two-dimensional, finite-element code are compared with the experimental results. These calculations reproduce the overall features of the flows but not the details of the stratified interface. The development of turbulence models for stratified flows is seen as a fruitful area for further research. (author)

  5. Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

    1980-09-01

    A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

  6. Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

    2016-11-01

    Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

  7. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    Science.gov (United States)

    Enoeda, M.; Kosaku, Y.; Hatano, T.; Kuroda, T.; Miki, N.; Honma, T.; Akiba, M.; Konishi, S.; Nakamura, H.; Kawamura, Y.; Sato, S.; Furuya, K.; Asaoka, Y.; Okano, K.

    2003-12-01

    This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  8. Operational-safety advantages of LMFBR's: the EBR-II experience and testing program

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lindsay, R.W.; Golden, G.H.

    1982-01-01

    LMFBR's contain many inherent characteristics that simplify control and improve operating safety and reliability. The EBR-II design is such that good advantage was taken of these characteristics, resulting in a vary favorable operating history and allowing for a program of off-normal testing to further demonstrate the safe response of LMFBR's to upsets. The experience already gained, and that expected from the future testing program, will contribute to further development of design and safety criteria for LMFBR's. Inherently safe characteristics are emphasized and include natural convective flow for decay heat removal, minimal need for emergency power and a large negative reactivity feedback coefficient. These characteristics at EBR-II allow for ready application of computer diagnosis and control to demonstrate their effectiveness in response to simulated plant accidents. This latter testing objective is an important part in improvements in the man-machine interface

  9. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  10. Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

    International Nuclear Information System (INIS)

    Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

    1979-09-01

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO 2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li 2 O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  11. Thermomechanical analysis of the DFLL test blanket module for ITER

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

    The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

  12. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Veleckis, E.; Yonco, R.M.; Maroni, V.A.

    1980-01-01

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li 3 N, Li 2 O, and Li 2 C 2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

  13. Neutronic Parametric Study on a Conceptual Design for a Transmutation Fusion Blanket

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2011-01-01

    Fusion energy may be the one of options of future energy. In all over the world, researchers are putting their efforts for its commercial and economical availability. Fusion-fission hybrid reactors have been studied for various applications in China. First milestone of fusion energy is expected to be the fusion fission hybrid reactors. In fusion-fission hybrid reactor the blanket design is of second prime importance after fusion source. In this study conceptual design of a fusion blanket is initiated for calculation of tritium production, transmutation of minor actinides (MA) and fission products (FP) and energy multiplication calculations

  14. Assessment of inspectability of LMFBR designs. Final report

    International Nuclear Information System (INIS)

    1981-09-01

    This two-volume report provides a comprehensive review of the inspectability of specific portions of loop- and pool-type LMFBR (1000-MWe) designs selected by EPRI. The designs were developed during the mid to late 1970s by three independent design teams (General Electric Co., Rockwell International, and Westinghouse) under the sponsorship of DOE (formerly ERDA) and EPRI. The requirements for normal, contingency, and post-repair inspections, addressed in this report, were established from Draft 12 of the ASME Boiler and Pressure Vessel Code, Section XI Division 3, issued in September 1979. These requirements, the intrinsic characteristics of the designs, the environmental (radiation, thermal, and atmospheric) aspects, and the available (present and near-term) inspection techniques, formed the basis for assessing the selected portions of the design or (1) accessibility, (2) feasibility, (3) practicality, and (4) costs to perform the above-specified inspections. Changes and additions fly ash has been as a concrete additive; however, extensive pilot scale development is underway to advance ash use in the TVA region in such areas as mineral and magnetite recovery, and mineral wool insulation. Recommended studies include: (1) the feasibility of converting existing wet fly d by the fuels include: residential (which includes residential and commercial), elthodology will be developed and verified in Phase II

  15. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

    Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

  16. Structural analysis for LMFBR applications[Indian position paper

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-05-01

    Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start.

  17. Experience on detection of leakages in LMFBR-steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Smit, C C

    1975-07-01

    One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks.

  18. Hydrodynamic analysis of the LMFBR prompt burst excursion (PBE) experiment

    International Nuclear Information System (INIS)

    Young, M.F.

    1977-01-01

    A series of in-pile experiments has been conducted at Sandia Laboratories to provide information on pressure levels and conversion of thermal energy into mechanical work in LMFBR cores during hypothetical, superprompt-critical excursions. Pressures generated in these experiments are recorded by a pressure transducer located at the top and bottom of a sodium channel surrounding a single, fresh UO 2 fuel pin. Work energy conversion is measured by a linear motion transducer connected to a piston at the top of the sodium column. Since the pressure transducers are located fairly far from the location of pin failure, it becomes necessary to determine the effect of channel geometry and piston motion on the observed pressure data. A two-dimensional, hydrodynamic analysis of pressure pulse propagation in the fuel pin-coolant channel geometry was therefore performed using the CSQII computer code. The initial series of PBE experiments consists of single, fresh UO 2 pins surrounded by a sodium-filled or dry-coolant channel contained in a closed test capsule. The capsule is subjected to a maximum pulse in the Annular Core Pulse Reactor (ACPR) resulting in an energy deposition of from 2350 to 2900 J/g (14 and 20 percent enriched pins). The pulse width at half maximum (PWHM) is about 5 ms

  19. Evaluation of high-pressure containment buildings for LMFBR's

    International Nuclear Information System (INIS)

    Armstrong, G.R.

    1981-01-01

    A study was conducted on the use of High Pressure LMFBR Containment Buildings for 1000 MW(e) LMFBRs. Two principal aspects were investigated: accident consequence mitigation and cost. Two types of hypothetical accidents were analyzed to establish consequence mitigation: melt-through and energetic expulsion. Three Containment Building (CB) design pressures were investigated: 69 kPa (10 psig), 207 kPa (30 psig), and 414 kPa (60 psig). Four types of design structures were analyzed to establish cost: steel, steel with confinement building, reinforced concrete, and prestressed/post-tensioned concrete. Results show that: it is within reason that a high pressure containment for a 1000 MW(e) reactor can be fabricated that will retain its integrity during postulated severe hypothetical accidents, if available measures are taken to reduce or prevent hydrogen production and the cost differential between basic high (414 kPa) and low (69 kPa) pressure containments is $10 x 10 6 or less

  20. Experience on detection of leakages in LMFBR-steam generators

    International Nuclear Information System (INIS)

    Smit, C.C.

    1975-01-01

    One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks

  1. Residual stress effects in LMFBR fracture assessment procedures

    International Nuclear Information System (INIS)

    Hooton, D.G.

    1984-01-01

    Two post-yield fracture mechanics methods, which have been developed into fully detailed failure assessment procedures for ferritic structures, have been reviewed from the point of view of the manner in which as-welded residual stress effects are incorporated, and comparisons then made with finite element and theoretical models of centre-cracked plates containing residual/thermal stresses in the form of crack-driving force curves. Applying the procedures to austenitic structures, comparisons are made in terms of failure assessment curves and it is recommended that the preferred method for the prediction of critical crack sizes in LMFBR austenitic structures containing as-welded residual stresses is the CEGB-R6 procedure based on a flow stress defined at 3% strain in the parent plate. When the prediction of failure loads in such structures is required, it is suggested that the CEGB-R6 procedure be used with residual/thermal stresses factored to give a maximum total stress of flow stress magnitude

  2. Multicell slug flow heat transfer analysis of finite LMFBR bundles

    International Nuclear Information System (INIS)

    Yeung, M.K.; Wolf, L.

    1978-12-01

    An analytical two-dimensional, multi-region, multi-cell technique has been developed for the thermal analysis of LMFBR rod bundles. Local temperature fields of various unit cells were obtained for 7, 19, and 37-rod bundles of different geometries and power distributions. The validity of the technique has been verified by its excellent agreement with the THTB calculational result. By comparing the calculated fully-developed circumferential clad temperature distribution with those of the experimental measurements, an axial correction factor has been derived to account for the entrance effect for practical considerations. Moreover, the knowledge of the local temperature field of the rod bundle leads to the determination of the effective mixing lengths L/sub ij/ for adjacent subchannels of various geometries. It was shown that the implementation of the accurately determined L/sub ij/ into COBRA-IIIC calculations has fairly significant effects on intersubchannel mixing. In addition, a scheme has been proposed to couple the 2-D distributed and lumped parameter calculation by COBRA-IIIC such that the entrance effect can be implanted into the distributed parameter analysis. The technique has demonstrated its applicability for a 7-rod bundle and the results of calculation were compared to those of three-dimensional analyses and experimental measurements

  3. LMFBR plant design features for sodium spill and fire protection

    International Nuclear Information System (INIS)

    Palm, R.E.

    1982-01-01

    Design features have been developed for an LMFBR plant to protect the concrete structures from potential liquid spills and fires and prevent sodium-concrete reactions. The inclusion of these features in the plant design reduces the severity of design basis accident conditions imposed on containment and other critical plant structures. Steel liners are provided in cells containing radioactive sodium systems, and catch pans are located in non-radioactive sodium system cells. The design requirements and descriptions of each of these protective features are presented. The loading conditions, analytical approach and numerical results are also included. Design of concrete cell structures that are subject to high temperature effects from sodium spills is discussed. The structural design considers the influence of high temperature on design properties of concrete and carbon steel materials based on results of a comprehensive test program. The development of these design features and high temperature design considerations for the Clinch River Breeder Reactor Plant (CRBRP) are presented in this paper

  4. Technical assessment study on pool-type LMFBR

    International Nuclear Information System (INIS)

    1986-01-01

    Technical assessment study on pool-type LMFBR was started in 1984 FY, inheriting the products from the Feasibility study, in order to accomplish cost reduction of reactor structure and enhanced structural reliability. This study consists of four major subjects; aseismic design development, component design optimization, high temperature structural design optimization and thermal hydraulics design optimization. In 1985 FY numbers of large model tests and analytical evaluations have been performed based on the prospects obtained in the first year's study. These tests and analyses have produced a lot of findings in each subject. They are concerning; (1) the effect of various building structures and analysis methods on floor response reduction, and data for evaluation of aseismic design concepts and structural integrity to seismic loading in the aseismic design development study. (2) data for evaluation of size reduction of main components in the reactor vessel, and heat transfer data required for structural integrity evaluation in the component design optimization study. (3) data for verification of inelastic analysis method, and assurance of technical applicability of disimilar weld in the high temperature structural design optimization study. (4) the effect of component size and location on thermal hydraulic characteristics, and data of thermal hydraulic similarity in thermal hydraulic design optimization study. This report summarizes the results obtained in 1985 FY. (author)

  5. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  6. Irradiation effects on low-friction coatings for LMFBR applications

    International Nuclear Information System (INIS)

    Ward, A.L.; Johnson, R.N.; Guthrie, G.L.; Aungst, R.C.

    1975-11-01

    A variety of wear-resistant low-friction materials has been irradiated in the EBR-II in order to assess their reponse to LMFBR environments. Pre- and postirradiation testing and examination efforts have concentrated on candidate materials for application to the wear pads on FTR ducts (fuel, control, and reflector assemblies), and a significant result has been qualification of a proprietary detonation-gun-applied chromium carbide coating which employs a Ni Cr binder. Additional materials such as Inconel-718, Haynes-273, aluminides, and various chromium carbide/binder combinations, and other application processes such as plasma-spray, weld-overlays, diffusion bonding and explosive bonding, have also been studied. The most detailed examinations were conducted on selected chromium carbide coatings and included visual inspection, weight and dimensional measurements, metallography, electron microprobe, epoxy-lift-off, and x-ray diffraction analysis. Chromium carbide coatings applied by the detonation-gun process have demonstrated a marked superiority to those applied by plasma-spray techniques

  7. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  8. Finite element elastic-plastic analysis of LMFBR components

    International Nuclear Information System (INIS)

    Levy, A.; Pifko, A.; Armen, H. Jr.

    1978-01-01

    The present effort involves the development of computationally efficient finite element methods for accurately predicting the isothermal elastic-plastic three-dimensional response of thick and thin shell structures subjected to mechanical and thermal loads. This work will be used as the basis for further development of analytical tools to be used to verify the structural integrity of liquid metal fast breeder reactor (LMFBR) components. The methods presented here have been implemented into the three-dimensional solid element module (HEX) of the Grumman PLANS finite element program. These methods include the use of optimal stress points as well as a variable number of stress points within an element. This allows monitoring the stress history at many points within an element and hence provides an accurate representation of the elastic-plastic boundary using a minimum number of degrees of freedom. Also included is an improved thermal stress analysis capability in which the temperature variation and corresponding thermal strain variation are represented by the same functional form as the displacement variation. Various problems are used to demonstrate these improved capabilities. (Auth.)

  9. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    International Nuclear Information System (INIS)

    Roake, W.E.

    1977-01-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

  10. Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

    International Nuclear Information System (INIS)

    Minden, C. v.; Schultheiss, G.F.

    1982-01-01

    A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.) [de

  11. Comparative analysis of LMFBR licensing in the United States and other countries - notably France. Executive summary

    International Nuclear Information System (INIS)

    Golay, M.W.; Castillo, M.

    1981-01-01

    The safety-related design aspects and licensing experiences of LMFBR projects in other democratic countries have been studied and contrasted to those in the United States in order to understand the importance of different approaches to safety, and also to understand better the system of the United States. The regulatory systems and LMFBR programs of France and the United States are contrasted in detail, and that of West Germany is also studied. The programs of Japan and the United Kingdom receive considerably less attention, and that of the Soviet Union is ignored

  12. Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

    International Nuclear Information System (INIS)

    Shin, Y.S.; Wambsganss, M.W.

    1975-05-01

    This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration

  13. Cover gas seals. 11 - FFTF-LMFBR seal-test program, January-March 1974

    International Nuclear Information System (INIS)

    Kurzeka, W.; Oliva, R.; Welch, F.

    1974-01-01

    The objectives of this program are to: (1) conduct static and dynamic tests to demonstrate or determine the mechanical performance of full-size (cross section) FFTF fuel transfer machine and reactor vessel head seals intended for use in a sodium vapor - inert gas environment, (2) demonstrate that these FFTF seals or new seal configuration provide acceptable fission product and cover gas retention capabilities at LMFBR Clinch River Plant operating environmental conditions other than radiation, and (3) develop improved seals and seal technology for the LMFBR Clinch River Plant to support the national objective to reduce all atmospheric contaminations to low levels

  14. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

    1977-04-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.

  15. LMFBR safety criteria: cost-benefit considerations under the constraint of an a priori risk criterion

    International Nuclear Information System (INIS)

    Hartung, J.

    1979-01-01

    The role of cost-benefit considerations and a priori risk criteria as determinants of Core Disruptive Accident (CDA)-related safety criteria for large LMFBR's is explored with the aid of quantitative risk and probabilistic analysis methods. A methodology is described which allows a large number of design and siting alternatives to be traded off against each other with the goal of minimizing energy generation costs subject to the constraint of both an a priori risk criterion and a cost-benefit criterion. Application of this methodology to a specific LMFBR design project is described and the results are discussed. 5 refs

  16. First-wall and blanket engineering development for magnetic-fusion reactors

    International Nuclear Information System (INIS)

    Baker, C.; Herman, H.; Maroni, V.; Turner, L.; Clemmer, R.; Finn, P.; Johnson, C.; Abdou, M.

    1981-01-01

    A number of programs in the USA concerned with materials and engineering development of the first wall and breeder blanket systems for magnetic-fusion power reactors are described. Argonne National Laboratory has the lead or coordinating role, with many major elements of the research and engineering tests carried out by a number of organizations including industry and other national laboratories

  17. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    Le Marois, G.; Federzoni, L.; Bucci, P.; Revirand, P.

    2000-01-01

    The benefit of HIP techniques applied to the fabrication of fusion internal components for higher performances, reliability and cost savings are emphasized. To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA. A coiled tube concept that allows cooling arrangement flexibility, strong reduction of the machining and number of welds is proposed for ITER IAM. Medium size mock-ups according to the WCLL breeding blanket concept have been manufactured. The fabrication of a large size mock-up is under progress. These activities are supported by numerical calculations to predict the deformations of the parts during HIP'ing. Finally, several HIP techniques issues have been identified and are discussed

  18. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

  19. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

    The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

  20. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Morley, N.B.

    1995-01-01

    Non-uniform flow distribution between parallel channels is one of the most serious concerns for self-cooled liquid metal blankets with electrically insulated walls. We show that uncertainties in flow distribution can be dramatically reduced by relatively simple design modifications. Several design features which impose flow uniformity by electrically coupling parallel channels are surveyed. Basic mechanisms for ''flow balancing'' are described, and a particular self-regulating concept using discrete passive electrodes is proposed for the US ITER advanced blanket concept. Scoping calculations suggest that this simple technique can be very powerful in equalizing the flow, even with massive insulator failures in individual channels. More detailed analyses and experimental verification will be required to demonstrate this concept for ITER. (orig.)

  1. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

  2. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  3. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

    Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

    2015-01-01

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  4. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  5. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m 2 and a particle heat flux of 1 MW/m 2 . Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  6. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

    Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

    2013-08-01

    ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

  7. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  8. Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    Science.gov (United States)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-02-01

    This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

  9. Key achievements in elementary R and D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    International Nuclear Information System (INIS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-01-01

    This paper presents the significant progress made in the research and development (R and D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li 2 TiO 3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 0 C followed by normalizing it at 930 0 C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R and D on the breeder material, Li 2 TiO 3 , the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li 2 TiO 3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li 2 TiO 3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation

  10. Key achievements in elementary R and Ds on water-cooled solid breeder blanket for ITER Test Blanket Module in JAERI

    International Nuclear Information System (INIS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Tanigawa, H.; Tobita, K.; Akiba, M.; Hayashi, K.; Ochiai, K.; Nishitani, T.

    2005-01-01

    This paper presents significant progress in research and development (R and D) of key elementary technologies on the water-cooled solid breeder blanket for the ITER test blanket modules (TBMs) in JAERI. Development of module fabrication technology, bonding technology of armors, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup, and tritium release behavior from Li 2 TiO 3 pebble bed under neutron pulsed operation condition are summarized. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 deg C followed by normalizing at 930 deg C after the Hot Isostatic Pressing (HIP) process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a solid state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it was found that the thermal fatigue lifetime of F82H can be predicted by using Manson-Coffin's law. As for R and Ds on a breeder material, Li 2 TiO 3 , effective thermal conductivity of Li 2 TiO 3 pebble was measured under compressive force simulating the ITER TBM environment. The increase in the effective thermal conductivity of the pebble bed was about 2.5 % at the compressive strain of 0.9 % at 400 deg C. Neutronic performance of the blanket module mockup has been carried out by the 14 MeV neutron irradiation. It was confirmed that the measured tritium production rate agreed with the calculated values within about 10% difference. Also, tritium release from a Li 2 TiO 3 pebble bed was measured under pulsed neutron irradiation conditions simulating the ITER operation. (author)

  11. Dynamic response of single hexagonal LMFBR core subassembly wrappers

    Energy Technology Data Exchange (ETDEWEB)

    Ash, J. E.; Marciniak, T. J.; (Argonne National Lab., IL (United States))

    1977-07-01

    To analyze the dynamic structural response of the LMFBR core subassembly hexagonal wrappers to postulated local energy releases and the sensitivity of the response to variations in both the pressure loading and the material properties of the stainless steel, a finite-element computer code STRAW has been developed. A series of experiments was performed to study the effects of variations in material properties. The amount of coldworking to which the Type 316 stainless steel is subjected has a strong influence upon the ductility and the elastic yield point. The usual fabrication process produced a nominally 20% coldworking with a yield point of about 680 MPa. By designing a special set of dies for the drawing process, a very low ductility hexcan was produced for which the yield point was raised to 820 MPa. Conversely, the yield point was lowered to 170 MPa by a solution annealing process producing a highly ductile test hexcan. A metallurgical study was conducted to find a representative brittle simulant material for the irradiated end-of-life steel properties. An aging treatment for Type 446 stainless steel was developed which reproduced the expected tensile-flow behavior of the in-pile subassembly. Further study is underway to investigate the fracture properties of the simulant material. The pressure pulses were generated by the controlled expansion of high-pressure detonation poducts from low-density explosives detonated inside a vented steel cannister. The orifice configuration of the cannister and the charge mixture ratio were designed to produce two specified pulse shapes. A charge containing 37,7 g PETN mixed with 35 wt % inert, hollow-glass microballoons developed a pressure pulse peak of 9.5 MPa at 1.0 ms. Increasing the PETN to 41 g resulted in a 14.6 MPa peak pressure, and increasing the explosive concentration to 90 wt % in the mixture increased the burning rate and the pulse risetime, so that the peak occurred at 0.6 ms.

  12. Fuel pin response to an overpower transient in an LMFBR

    International Nuclear Information System (INIS)

    Grosberg, A.J.; Head, J.L.

    1979-01-01

    This paper describes a method by which the ability of a whole-core code accurately to predict the time and location of the first fuel pin failures may be tested. The method involves the use of a relatively simple whole-core code to 'drive' a sophisticated fuel pin code, which is far too complex to be used within a whole-core code but which is potentially capable of modelling reliably the response of an individual fuel pin. The method cannot follow accurately the subsequent course of the transient because the simple whole-core code does not model the reactivity effects of events which may follow pin failure. The codes used were the simple whole-core code FUTURE and the fuel pin behaviour code FRUMP. The paper describes an application of the method to analyse a hypothetical LMFBR accident in which the control rods were assumed to be driven from the core at maximum speed, with all trip circuits failed. Taking 0.5% clad strain as a clad failure criterion, failure was predicted to occur at the top of the active core at about 10s into the transient. A repeat analysis, using an alternative clad yield criterion which is thought to be more realistic, indicated failure at the same position but 24s into the transient. This is after the onset of sodium boiling. Pin failure at the top of the core are likely to cause negative reactivity changes. In this hypothetical accident, pin failures are likely, therefore, to have a moderating effect on the course of the transient. (orig.)

  13. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  14. Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

  15. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

    Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

  16. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  17. Input parameters to codes which analyze LMFBR wire-wrapped bundles

    International Nuclear Information System (INIS)

    Hawley, J.T.; Chan, Y.N.; Todreas, N.E.

    1980-12-01

    This report provides a current summary of recommended values of key input parameters required by ENERGY code analysis of LMFBR wire wrapped bundles. This data is based on the interpretation of experimental results from the MIT and other available laboratory programs

  18. The state of art of the methods for thermohydraulics design of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1981-09-01

    The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt

  19. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  20. LARA: Expert system for acoustic localization of robot in a LMFBR

    International Nuclear Information System (INIS)

    Lhuillier, C.; Malvache, P.

    1986-12-01

    The expert system LARA (Acoustic Localization of Autonomic Robot) has been developed to show the interest of introducing artificial intelligency for fine automatic positioning of refuelling machine in a LMFBR reactor. LARA which is equipped with an acoustic detector gives rapidly a good positioning on the fuel [fr

  1. Subassembly faults diagnostic of an LMFBR type reactor by the measurement of temperature noise

    International Nuclear Information System (INIS)

    Kokorev, B.V.; Palkin, I.I.; Turchin, N.M.; Pallagi, D.; Horanyi, S.

    1979-09-01

    The subassembly faults detection possibility by temperature noise analysis of an LMFBR is described. The paper contains the results of diagnostical examinations obtained on electrically heated NaK test rigs. On the basis of these results the measurement of temperature noise RMS value seems to be a practicable method to detect local blockages in an early phase. (author)

  2. LMFBR safety program. Annual technical progress report. Government fiscal year, 1977

    International Nuclear Information System (INIS)

    1977-01-01

    Information is presented concerning the development of the SOMIX-1 computer code for sodium drop burning analysis; experimental analysis of burning sodium drops; aerosol leakage from containment buildings; high-temperature-concentration aerosols; aerosol source term from vaporized fuel; properties of high-temperature fuel mixtures; and development of the COMRADEX computer code for analysis of radiological doses in the environment from LMFBR accidents

  3. Review of pertinent thermal-hydraulic data for LMFBR core natural circulation analyses

    International Nuclear Information System (INIS)

    Bishop, A.A.; Coffield, R.D. Jr.; Markley, R.A.

    1980-01-01

    A literature review and summary of significant data is presented relative to LMFBR core natural convection cooling analysis. First, a brief review of computer codes and respective input data needs is made, significant data areas are then addressed and data for verifying the code calculations are described. Recommendations and conclusions with regard to the data are included

  4. A survey of the French creep-fatigue design rules for LMFBR

    International Nuclear Information System (INIS)

    Tribout, J.; Cordier, G.; Moulin, D.

    1987-01-01

    The paper provides a survey of the creep-fatigue design rules for the LMFBR in France. These rules are the ones currently implemented in French component manufacturing. The background of each item is discussed and the trends for improvements currently investigated are described. The creep-fatigue rules apply to elastic analysis only. (orig.)

  5. A miniature inductive temperature sensor to monitor temperature noise in the coolant of an LMFBR

    International Nuclear Information System (INIS)

    Dean, S.A.; Sandham, C.W.

    1980-01-01

    A description is given of the design and performance of miniature inductive sensors developed to monitor fast temperature fluctuations in the sodium coolant above the core of a LMFBR. These instruments, designed to be installed within existing thermocouple containment thimbles, also provide a steady-state temperature indication for reactor control purposes. (author)

  6. Theoretical study and experimental investigation of mixed and natural circulation in LMFBR core subassemblies

    International Nuclear Information System (INIS)

    Leteinturier, D.; Blanc, D.; Menant, B.; Basque, G.

    1980-02-01

    A presentation is made of theoretical and experimental studies carried out in France on mixed and natural convection in LMFBR wire wrapped bundles. Two codes are described, one for mixed convection THERNAT and the other for natural convection BACCHUS. THe related experimental program FETUNA, with electrically heated bundles in sodium loops, is also presented

  7. The water vapor nitrogen process for removing sodium from LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    Crippen, M D; Funk, C W; Lutton, J M [Hanford Engineering Development Laboratory, Richland (United States)

    1978-08-01

    Application and operation of the Water Vapor-Nitrogen Process for removing sodium from LMFBR components is reviewed. Emphasis is placed on recent efforts to verify the technological bases of the process, to refine the values of process parameters and to ensure the utility of the process for cleaning and requalifying components. (author)

  8. Contribution of the CEC in structural analysis applied to LMFBR problems

    International Nuclear Information System (INIS)

    Larsson, L.H.; Terzaghi, A.

    1983-01-01

    This paper presents both the activity of DG XII in field of Codes and Standards (harmonization) and the research activity carried out at the JRC in Ispra. The first part describes the activity performed in the field of structural analysis by the Fast Reactor Coordinating Committee of the CEC and its Working Group Codes and Standards. This activity, which is aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, has made good progress in most areas. Results from recent inelastic and seismic benchmark calculations are presented as well as future computational exercises and investigations related to piping analysis, defect analysis, material behaviour and life prediction at elevated temperature. In the second part of the paper results of recent research and future plans in the area of structural mechanics at the JRC Ispra are discussed. In the past years, a large effort was devoted to the COVA (code validation) program intended to validate dynamic fluid/structure codes necessary for predicting the response of LMFBR containments. The main conclusions that can be drawn from COVA which finishes this year are presented, and some still open questions related to the prediction of containment response to an HCDA are discussed. The paper then describes the identification technique which is applied for the determination of constitutive equations for the dynamic behaviour of materials. In the field of fracture mechanics JRC has mostly concentrated its efforts on the elastic-plastic fracture toughness properties of irradiated austenitic steels. In the future, also dynamic ductile fracture problems will be investigated, for these a large dynamic test facility with a max. force of 5 MN will be used. The numerical analysis methods associated with these tests are discussed. (author)

  9. Safety issues for LMFBR: important features drawn from the assessments of Superphenix

    International Nuclear Information System (INIS)

    Natta, M.

    2002-01-01

    Superphenix, which is built on the site of Creys-Malville, is still the biggest LMFBR plant that has been in operation. It is a pool type reactor, as Phenix and the RNR 1 500 and EFR projects. After the analysis of the preliminary safety (1974-1975), the construction was authorised by decree of the Prime Minister in 1977, the authorization for fuel loading and star-up to 3% was given by the minister of industry in July 1985 and full power was achieved in December 1986. The plant was operated until the end of December 1996, producing the equivalent of 320 EFPD, corresponding to half of the maximum barn-up of the first core. The plant was definitively stopped on the 20. of April 1998 by a decision of the French government. During this period of 25 years of licensing, construction and operation of Superphenix, others discussions and preliminary licensing procedures were started for new projects, mainly the RNR 1500 French project and the EFR European project. The operation of Superphenix was also marked by several incidents, which led to additional licensing procedures and important modifications. This period was also marked by an important work of research and development in the safety field, mostly related to the issues concerning hypothetical core disruptive accidents (HCDA) and sodium fires; further, this period was marked by the Three Mile Island accident in 1979 and the Chernobyl accident in 1986. The purpose of this paper is to present some items which were discussed during this period of 25 years and which should be of interest for future LMFBRs. In this presentation, we shall discuss the key issues concerning the safety criteria and options taken with respect to severe accidents, i.e. core melt accidents, giving details on some specific which are less known since they were assessed only lately for Superphenix, sometimes in connection with the on-going safety researches. (author)

  10. Processing and waste disposal representative for fusion breeder blanket systems

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1987-01-01

    This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made

  11. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    Jackson, D.P.; Selander, W.N.; Townes, B.M.

    1985-01-01

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  12. Hardware concepts for a large low-energetics LMFBR core. Final report

    International Nuclear Information System (INIS)

    Hutter, E.; Batch, R.V.

    1980-12-01

    A design study was made to identify a practical set of hardware configurations that would embody the requirements developed in the numerical study of a low-energetics core and blanket for a prototype large breeder reactor. Dimensioned drawings are presented for fuel, blanket, reflector/shield, and control rod subassemblies. A horizontal cross section drawing shows how these subassemblies are arranged in the total core/blanket assembly. A core support is illustrated showing a dual plenums arrangement

  13. Electromagnetic analysis of ITER shield blanket under VDE

    International Nuclear Information System (INIS)

    Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

    2010-01-01

    Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

  14. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

    A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

  15. SIMMER-I: an S/sub n/, Implicit, Multifield, Multicomponent, Eulerian, Recriticality code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Bell, C.R.; Bleiweis, P.B.; Boudreau, J.E.; Parker, F.R.; Smith, L.L.

    1976-08-01

    Physical models, numerical methods, and program description are presented for SIMMER-I, a computer program which predicts the neutronic and fluid dynamic behavior of an LMFBR during a hypothetical core disruptive accident

  16. Complementary role of critical integral experiment and power reactor start-up experiments for LMFBR neutronics data and method validation

    International Nuclear Information System (INIS)

    Salvatores, M.

    1986-09-01

    Both critical experiments and power reactor results play at present a complementary role in reducing the uncertainties in Key design parameters for LMFBR, which can be relevant for the economic performances of this type of reactors

  17. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

  18. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.; Smith, D.L.

    1984-01-01

    Consideration of the required functions of the blanket and the general chemical, mechanical, and physical properties of candidate tritium breeding materials, coolants, structural materials, etc., leads to acceptable or compatible combinations of materials. The presently favored candidate structural materials are the austenitic stainless steels, martensitic steels, and vanadium alloys. The characteristics of these alloy systems which limit their application and potential performance as well as approaches to alloy development aimed at improving performance (temperature capability and lifetime) will be described. Progress towards understanding and improving the performance of structural materials has been substantial. It is possible to develop materials with acceptable properties for fusion applications

  19. About possible technologies of creation nanostructures blankets

    International Nuclear Information System (INIS)

    Blednova, Zh.M.; Chaevskij, M.I.; Rusinov, P.O.

    2008-01-01

    Possible technologies of formation nanostructures blankets are considered: a method of thermal carrying over of weights in the conditions of a high gradient of temperatures; the combined method including cathode-plasma nitriding in the conditions of low pressure and drawing of nitride of the titan in a uniform work cycle; the combined method including high-frequency ionic nitriding and drawing of carbide of chrome by pyrolysis chrome and organic of connections in plasma of the decaying category. Possibility of formation layered nanostructures layers is shown.

  20. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-1 through 4 and PULSAR 1 and 2. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. Also, the requirements of engineering and physics systems for a pulsed reactor were evaluated by the PULSAR design studies. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies

  1. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    Noda, Kenji

    1998-03-01

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li 2 TiO 3 , tritium release behavior of Li 2 TiO 3 and Li 2 ZrO 3 including tritium diffusion, modeling of tritium release from Li 2 ZrO 3 in ITER condition, helium release behavior from Li 2 O, results of tritium release irradiation tests of Li 4 SiO 4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  2. Welding techniques development of CLAM steel for Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Li Chunjing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China)], E-mail: lcj@ipp.ac.cn; Huang Qunying; Wu Qingsheng; Liu Shaojun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Lei Yucheng [Jiangsu University, Zhenjiang, Jiangsu, 212013 (China); Muroga, Takeo; Nagasaka, Takuya [National Institute for Fusion Science, Toki, Jifu, 509-5292 (Japan); Zhang Jianxun [Xi' an Jiaotong University, Xi' an, Shanxi, 710049 (China); Li Jinglong [Northwestern Polytechnical University, Xi' an, Shanxi, 710072 (China)

    2009-06-15

    Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.

  3. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  4. Two-phase-flow cooling concept for fusion reactor blankets

    International Nuclear Information System (INIS)

    Bender, D.J.; Hoffman, M.A.

    1977-01-01

    The new two-phase heat transfer medium proposed is a mixture of potassium droplets and helium which permits blanket operation at hih temperature and low pressure, while maintaining acceptable pumping power requirements, coolant ducting size, and blanket structure fractions. A two-phase flow model is described. The helium pumping power and the primary heat transfer loop are discussed

  5. Overview of the TFTB lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The Lithium Blanket Module (LBM) is an ∼ 80-cm 3 module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program

  6. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

    Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

    2001-01-01

    In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R and D program, a rail-mounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R and D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance. (author)

  7. Accelerator driven heavy water blanket on circulating fuel

    International Nuclear Information System (INIS)

    Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

    1997-01-01

    A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

  8. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-01-01

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored

  9. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

    Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

    2016-11-01

    Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

  10. 18 CFR 284.303 - OCS blanket certificates.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false OCS blanket certificates. 284.303 Section 284.303 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... Pipelines on Behalf of Others § 284.303 OCS blanket certificates. Every OCS pipeline [as that term is...

  11. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bertone, P.C.; Creedon, R.L.; File, J.; Graumann, D.W.

    1985-01-01

    The Lithium Blanket Module (LBM) is an approximately 80X80X80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li 2 O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li 2 O pellets with satisfactory reproducibility were developed using purified Li 2 O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1-1nCi/g)

  12. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

    Cheng, E.T.; Engholm, B.A.; Su, S.D.

    1981-01-01

    The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

  13. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

  14. Neutron dosimetry for the TFTR Lithium-Blanket-Module program

    International Nuclear Information System (INIS)

    Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

    1981-01-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

  15. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Barleon, L.

    1988-01-01

    A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

  16. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  17. EU contribution to the procurement of the ITER blanket first wall

    International Nuclear Information System (INIS)

    Lorenzetto, Patrick; Banetta, Stefano; Bellin, Boris; Boireau, Bruno; Bucci, Philippe; Cicero, Tindaro; Conchon, Denis; Dellopoulos, Georges; Hardaker, Stephen; Marshall, Paul; Nogué, Patrice; Pérez, Marcos; Gutierrez, Leticia Ruiz; Samaniego, Fernando; Sherlock, Paul; Zacchia, Francesco

    2016-01-01

    Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

  18. EU contribution to the procurement of the ITER blanket first wall

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, Patrick, E-mail: Patrick.Lorenzetto@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Banetta, Stefano; Bellin, Boris [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Boireau, Bruno [AREVA NP, Centre Technique, 71200 Le Creusot (France); Bucci, Philippe [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Cicero, Tindaro [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Conchon, Denis [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Dellopoulos, Georges [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Hardaker, Stephen [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Marshall, Paul [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Nogué, Patrice [AREVA NP, Centre Technique, 71200 Le Creusot (France); Pérez, Marcos [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Gutierrez, Leticia Ruiz [Iberdrola Ingeniería y Construcción S.A.U., Avenida Manoteras 20, 28050 Madrid (Spain); Samaniego, Fernando [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Sherlock, Paul [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Zacchia, Francesco [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain)

    2016-11-01

    Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

  19. Assessing impact of blanket interventions for MAM prevention

    International Nuclear Information System (INIS)

    Grais, Rebecca F.; Isanaka, I; Langendorf, C; Roederer, T

    2014-01-01

    Full text: Blanket interventions for MAM prevention (Blanket supplementary feeding programming (BSFP)) provide a supplementary food ration often accompanied by a basic medical treatment and prevention package to a vulnerable population for a defined period in a defined geographic location. There is little strong evidence on the impact of BSFP on rates of malnutrition and mortality, and scare guidance on program monitoring and evaluation to improve the implementation of specific programs. Assessing the impact of BSFP has been fraught with difficulty. Their isolated impact is difficult, if not often impossible to disentangle from larger care and prevention packages, the objectives of BSFP may vary by context, implementing agency, time and geography. Various and often multiple co-morbidities among children in the targeted group complicate matters further with respect to impact assessment. This leads to difficulties in generalizing results from one context to another and the need for more complex metrics to guide operational decision-making. Ideally, impact or effectiveness of BSFP should be addressed in a research framework where appropriate and complete data is collected in order to address specific questions. The gold standard is the conduct of randomized studies including a control group. These studies have been scarce as they may be perceived as either rarely feasible or not ethical or both. However, as generating evidence on impact of BSFP is essential to provide operational guidance, these studies should be encouraged through a diversity of robust, yet creative and pragmatic, methodological approaches. As a case study, a series of studies conducted over the past decade are reviewed in the same location in Niger highlighting the lessons learned. (author)

  20. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1984-08-01

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6 Li in order to reduce parasite neutron captures in there. (orig./HP)

  1. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  2. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    International Nuclear Information System (INIS)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li 2 O) and lithium zirconate (Li 2 ZrO 3 ) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

  3. Evaluation of air cleaning system concepts for emergency use in LMFBR plants

    International Nuclear Information System (INIS)

    Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

    1976-12-01

    Nineteen different air cleaning concepts are arranged into twenty-four systems and evaluated for use as accident mitigating systems in LMFBR plants. Both single, low-leakage containment plants and once-through operation applicable to containment/confinement plants are considered. Plant characteristics affecting air cleaning requirements are defined for 1000 MW(e) plants and a sodium and radiological release term is postulated. The accident conditions under which the emergency air cleaning system (EACS) must function is established by use of SOFIRE-II and HAA-3B computer codes. Criteria are developed for evaluating the various systems and for assigning comparative ratings. The numerical ratings are combined with information on cost and development potential to arrive at recommendations for the most promising systems. The conclusion is made that reliable and effective systems are feasible for use as engineered safety features for LMFBR plants, but that development effort is required for all the air cleaning concepts evaluated

  4. Single-phase sodium pump model for LMFBR thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.

    1979-01-01

    A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics

  5. LMFBR operational and experimental local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS- and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  6. A critical experimental study of integral physics parameters in simulated LMFBR meltdown cores

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Wade, D.C.; Bucher, R.G.; Smith, D.M.; McKnight, R.D.; Lesage, L.G.

    1978-01-01

    Integral physics parameters of several representative, idealized meltdown LMFBR configurations were measured in mockup critical assemblies on the ZPR-9 reactor at Argonne National Laboratory. The experiments were designed to provide data for the validation of analytical methods used in the neutronics part of LMFBR accident analysis. Large core distortions were introduced in these experiments (involving 18.5% core volume) and the reactivity worths of configuration changes were determined. The neutronics parameters measured in the various configurations showed large changes upon core distortion. Both diffusion theory and transport theory methods were shown to mispredict the experimental configuration eigenvalues. In addition, diffusion theory methods were shown to result in a non-conservative misprediction of the experimental configuration change worths. (author)

  7. LMFBR operational and experimental in-core local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS-and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  8. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle

  9. Acoustics and voiding dynamics during SLSF simulations of LMFBR undercooling transients

    International Nuclear Information System (INIS)

    Anderson, T.T.; Kuzay, T.M.; Marr, W.W.; Miles, K.J.; Pedersen, D.R.; Thompson, D.H.; Wilson, R.E.

    1978-01-01

    The SLSF is the largest U.S. in-reactor test vehicle for steady-state and transient experiments in an environment typical of a LMFBR core. The SLSF experiment program, sponsored by the Department of Energy, contributes to the LMFBR safety assurance program by providing data on key phenomena that occur during postulated reactor accidents. This paper describes completed SLSF experiments, in-core instrumentation used, and methods of data interpretation to determine sodium boiling and voiding dynamics. Boiling inception is shown to be identifiable from several types of in-core instruments. Location of the boiling front and void growth derived from experimental data are compared with analytical predictions. These and other data form the basis to improve understanding of accidents and to validate or guide the development of accident analysis methods

  10. Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-10-01

    The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors.

  11. Report of the IAEA advisory group meeting on LMFBR fuel reprocessing

    International Nuclear Information System (INIS)

    1976-05-01

    A summary of the papers and discussions of the meeting is presented, reviewing the status of development in LMFBR fuel reprocessing and focusing attention on important problem areas. The following topics are discussed: Transport, storage and removal of sodium; decladding and shearing; dissolution; Purex process; fluoride volatility method; off-gas purification; waste disposal. Status reports of national programmes of Belgium, France, Federal Republic of Germany, Italy, Japan, United Kingdom, USSR and USA are included

  12. LMFBR conceptual design study: an overview of environmental and safety concerns

    International Nuclear Information System (INIS)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE

  13. A survey of LMFBR cavitation technology in the U.S.A

    International Nuclear Information System (INIS)

    Cha, Y.S.; Huebotter, P.R.; Hopenfeld, J.

    1976-01-01

    Several experimental programmes of a basic and applied nature were established in the USA in order to develop guidelines to ensure design and operation of LMFBR hydraulic components free from cavitation and/or cavitation damage. As of March 1976, most of these experimental programs are still in progress. Each programme is briefly described. The available interium data are presented. References that are relevant are provided

  14. LMFBR flexible pipe joint development program. Annual technical progress report, government fiscal year 1977

    International Nuclear Information System (INIS)

    1978-01-01

    Currently, the ASME Boiler and Pressure Vessel Code does not allow the use of flexible pipe joints (bellows) in Section III, Class 1 reactor primary piping systems. Studies have shown that the primary piping loops of LMFBR's could be simplified by using these joints. This simplification translates directly into shorter primary piping runs and reduced costs for the primary piping system. Further cost savings result through reduced vault sizes and reduced containment building diameter. In addition, the use of flexible joints localizes the motions from thermally-induced piping growth into components which are specifically designed to accommodate this motion. This reduces the stress levels in the piping system and its components. It is thus economically and structurally important that flexible piping joints be available to the LMFBR designer. The overall objective of the Flexible Joint Program is to provide this availability. This will be accomplished through the development of ASME rules which allow the appropriate use of such joints in Section III, Class 1 piping systems and through the development and demonstration of construction methods which satisfy these rules. The rule development includes analytic and testing methodology formulations which will be supported by subscale bellows testing. The construction development and demonstration encompass the design, fabrication, and in-sodium testing of prototypical LMFBR plant-size flexible pipe joints which meet all ASME rule requirements. The satisfactory completion of these developmental goals will result in an approved flexible pipe joint design for the LMFBR. Progress is summarized in the following efforts undertaken during 1977 to accomplish these goals: (1) code case support, (2) engineering and design, (3) material development, (4) testing, and (5) manufacturing development

  15. LMFBR conceptual design study: an overview of environmental and safety concerns

    Energy Technology Data Exchange (ETDEWEB)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE.

  16. Comparison of different LMFBR primary containment codes applied to a Benchmark problem

    International Nuclear Information System (INIS)

    Benuzzi, A.

    1986-01-01

    The Cont Benchmark calculation exercise is a project sponsored by the Containment Loading and Response Group, a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee - CEC. A full-size typical Pool type LMFBR undergoing a postulated Core Disruptive Accident (CDA) has been defined by Belgonucleaire-Brussels under a study contract financed by the CEC and has been submitted to seven containment code calculations. The results of these calculations are presented and discussed in this paper

  17. Comparative study of heterogeneous and homogeneous LMFBR cores in some accident conditions

    International Nuclear Information System (INIS)

    Renard, A.; Evrard, G.

    1978-01-01

    An heterogeneous design and a homogeneous one of a LMFBR core with the same power and similar dimensions are compared from the safety point-of-view. The comparison is performed for several accident conditions, such as Loss-of-Flow and Transient Overpower, with the same failure criteria and model assumptions for both cores. Qualitative trends are deduced from the behaviour of the core designs in the investigated transient conditions. (author)

  18. Monte-Carlo validation of secondary sodium activation in a pool-type LMFBR

    International Nuclear Information System (INIS)

    Plamiotti, G.; Rado, V.; Salvatores, M.

    1980-09-01

    The secondary sodium activation in a pool-type LMFBR is the main parameter to be accurately evaluated in the shield design. In the present work a complete two dimensional description of the system, including core, shielding and sodium up to Heat Exchangers, is coupled to local Heat Exchanger Monte-Carlo calculations. This refined calculation is used to deduce a simplified method to take into account the coupling of radial propagation in the Heat Exchanger and its finite cylindrical structure

  19. Seismic criteria studies and analyses. Quarterly progress report No. 3. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    1975-01-03

    Information is presented concerning the extent to which vibratory motions at the subsurface foundation level might differ from motions at the ground surface and the effects of the various subsurface materials on the overall Clinch River Breeder Reactor site response; seismic analyses of LMFBR type reactors to establish analytical procedures for predicting structure stresses and deformations; and aspects of the current technology regarding the representation of energy losses in nuclear power plants as equivalent viscous damping.

  20. Comprehensive method of common-mode failure analysis for LMFBR safety systems

    International Nuclear Information System (INIS)

    Unione, A.J.; Ritzman, R.L.; Erdmann, R.C.

    1976-01-01

    A technique is demonstrated which allows the systematic treatment of common-mode failures of safety system performance. The technique uses log analysis in the form of fault and success trees to qualitatively assess the sources of common-mode failure and quantitatively estimate the contribution to the overall risk of system failure. The analysis is applied to the secondary control rod system of an early sized LMFBR

  1. Fretting and wear of stainless and ferritic steels in LMFBR steam generators

    International Nuclear Information System (INIS)

    Lewis, M.W.J.; Campbell, C.S.

    1981-01-01

    Steam generators for LMFBR's may be subject to both fretting wear as a result of flow-induced vibrations and to wear from larger amplitude sliding movements from thermal changes. Results of tests simulating the latter are given for stainless and ferritic steels. For the assessment of fretting wear damage, vibration assessments must be combined with data on specific wear rates. Test mechanisms used to study fretting in sodium covering impact, impact-slide and pure rubbing are described and results presented. (author)

  2. Acoustic detection for water/steam leak from a tube of LMFBR steam generator

    International Nuclear Information System (INIS)

    Sonoda, Masataka; Shindo, Yoshihisa

    1989-01-01

    Acoustic leak detector is useful for detecting more quickly intermediate leak than the existing hydrogen detector and is available for identification of leak location on the accident of water/steam leak from a tube of LMFBR steam generator. This paper presents the overview of HALD (High frequency Acoustics Leak Detection) system, which is more sensitive for leak detection and lower cost of equipment for identification of leak location than a low frequency type detector. (author)

  3. Seismic analysis of a large LMFBR with fluid-structure interactions

    International Nuclear Information System (INIS)

    Ma, D.C.

    1985-01-01

    The seismic analysis of a large LMFBR with many internal components and structures is presented. Both vertical and horizontal seismic excitations are considered. The important hydrodynamic phenomena such as fluid-structure interaction, sloshing, fluid coupling and fluid inertia effects are included in the analysis. The results of this study are discussed in detail. Information which is useful to the design of future reactions under seismic conditions is also given. 4 refs., 12 figs

  4. Development of an 85,000 gpm (19,303 m3/h) LMFBR primary pump

    International Nuclear Information System (INIS)

    Zerinvary, M.C.; Wagner, E.W.

    1984-01-01

    The development of an 85,000 gpm two-stage primary pump for liquid metal fast breeder reactor (LMFBR) applications is described. The design was supported by air and cavitation model testing of the hyraulics, and development and feature testing of the level control system and the adjustable frequency solid state power supply. Important fabrication and water test items are also discussed, along with some unique assembly tooling requirements

  5. Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    1978-10-01

    The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors

  6. LMFBR system-wide transient analysis: the state of the art and US validation needs

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.

    1982-01-01

    This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants

  7. A fundamental study on sodium-water reaction in the double-pool-type LMFBR

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Akimoto, Tokuzo

    1987-01-01

    In order to evaluate the pressure rise by large sodium-water reaction in the Double-Pool LMFBR, basic tests on pressure wave celerity in rectangular tube are carried out. The initial spike pressure in rectangular-shelled steam generator of the Double Pool reactor, strongly depends on pressure wave celerity. In this study, celerity was measured as a function of pressure wave rising time and pulse height, and influence of water around the test section on celerity was investigated. (author)

  8. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  9. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  10. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    Ishitsuka, E.

    2002-01-01

    Advanced solid breeding blanket design in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high dose of neutron irradiation. Therefore, the development of such advanced blanket materials is indispensable. In this paper, the cooperation activities among JAERI, universities and industries in Japan on the development of these advanced materials are reported. Advanced tritium breeding material to prevent the grain growth in high temperature had to be developed because the tritium release behavior degraded by the grain growth. As one of such materials, TiO 2 -doped Li 2 TiO 3 has been studied, and TiO 2 -doped Li 2 TiO 3 pebbles was successfully fabricated. For the advanced neutron multiplier, the beryllium intermetallic compounds that have high melting point and good chemical stability have been studied. Some characterization of Be 12 Ti was studied. The pebble fabrication study for Be 12 Ti was also performed and Be 12 Ti pebbles were successfully fabricated. From these activities, the bright prospect to realize the DEMO blanket by the application of TiO 2 -doped Li 2 TiO 3 and beryllium intermetallic compounds was obtained. (author)

  11. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations

  12. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed

  13. Neutronics analysis for aqueous self-cooled fusion reactor blankets

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Jaffa, R.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1986-06-01

    The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

  14. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

    1992-06-01

    Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

  15. Damping of the radial impulsive motion of LMFBR core components separated by fluid squeeze films

    International Nuclear Information System (INIS)

    Liebe, R.; Zehlein, H.

    1977-01-01

    The core deformation of a liquid metal cooled fast breeder reactor (LMFBR) due to local pressure propagation from rapid energy releases is a complex three-dimensional fluid-structure-interaction problem. High pressure transients of short duration cause structural deformation of the closely spaced fuel elements, which are surrounded by the flowing coolant. Corresponding relative displacements give rise to squeezing fluid motion in the thin layers between the subassemblies. Therefore significant backpressures are produced and the resulting time and space dependent fluid forces are acting on the structure as additional non-conservative external loads. Realistic LMFBR safety analysis of several clustered fuel elements have to account for such flow induced forces. Several idealized models have been proposed to study some aspects of the complex problem. As part of the core mechanics activities at GfK Karlsruhe this paper describes two fluid flow models (model A, model B), which are shown to be suitable for physically coupled fluid-structure analyses. Important assumptions are discussed in both cases and basic equations are derived for one- and two-dimensional incompressible flow fields. The interface of corresponing computer codes FLUF (model A) and FLOWAX (model B) with structural dynamics programs is outlined. Finally fluid-structure interaction problems relevant to LMFBR design are analyzed; parametric studies indicate a significant cushioning effect, energy dissipation and a strongly nonlinear as well as timedependent damping of the structural response. (Auth.)

  16. Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

    1976-01-01

    A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

  17. Overview of EU activities on DEMO liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

    1994-01-01

    The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

  18. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

  19. Molten salt cooling/17Li-83Pb breeding blanket concept

    International Nuclear Information System (INIS)

    Sze, D.K.; Cheng, E.T.

    1985-02-01

    A description of a fusion breeding blanket concept using draw salt coolant and static 17 Li- 83 Pb is presented. 17 Li- 83 Pb has high breeding capability and low tritium solubility. Draw salt operates at low pressure and is inert to water. Corrosion, MHD, and tritium containment problems associated with the MARS design are alleviated because of the use of a static LiPb blanket. Blanket tritium recovery is by permeation toward the plasma. A direct contact steam generator is proposed to eliminate some generic problems associated with a tube shell steam generator

  20. Conceptual design of blanket structures for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

  1. Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

    International Nuclear Information System (INIS)

    1978-01-01

    This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified

  2. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

    1989-01-01

    The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

  3. Cassette blanket and vacuum building: key elements in fusion reactor maintenance

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

    The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

  4. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

    2016-01-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  5. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

    2016-11-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  6. {open_quotes}Transmutation efficiency calculation in the blanket on melted salts with central neptunium target{close_quotes}

    Energy Technology Data Exchange (ETDEWEB)

    Kolesov, V.F.; Shtarev, S.K.; Khoruzhiy, V.K. [Russia Federal Nuclear Center, Arzamas (Russian Federation)] [and others

    1995-10-01

    In the limits of ABC project version of two-sectional reactor system in the form of combination of subcritical blanket on melted salts and multiplying target from threshold fissile material {sup 237}Np is considered. This research is the development of the VNIIEF`s earlier work`s (Russia) investigating of usage possibilities in ABC project the conception of multisectional blankets with single-sided neutron coupling between sections. With the help of Monte-Carlo program the calculations results of system mentioned are given. The possibility of accelerator`s considerable power reduction at the account of thorium target substitution with neptunium-237 multiplying target is shown.

  7. Evaluation of the LMFBR cover gas source term and synthesis of the associated R and D

    International Nuclear Information System (INIS)

    Balard, F.; Carluec, B.

    1996-01-01

    At the end of the seventies and the beginning of the eighties, there appeared a pressing need of experimental results to assess the LMFBR's safety level. Because of the urgency, analytical studies were not systematically undertaken and maximum credible cover gas instantaneous source terms (radionuclides core release fraction) were got directly from crude out-of-pile experiment interpretations. Two types of studies and mock-ups were undertaken depending on the timescale of the phenomena: instantaneous source terms (corresponding to an unlikely energetic core disruptive accident CDA), and delayed ones (tens of minutes to some hours). The experiments performed in this frame are reviewed in this presentation: 1) instantaneous source term: - FAUST experiments: I, Cs, UO2 source terms (FzK, Germany), - FAST experiments : pool depth influence on non volatile source term (USA), - CARAVELLE experiments: nonvolatile source term in SPX1 geometry (CEA, France); 2) delayed source term: - NALA experiments: I, Cs, Sr, UO2 source term (FzK, Germany), - PAVE experiments: I source term (CEA, France), - NACOWA experiments: cover gas aerosols enrichment in I and Cs (FzK, Germany) - other French experiments in COPACABANA and GULLIVER facilities. The volatile fission products release is tightly bound to sodium evaporation and a large part of the fission products is dissolved in the liquid sodium aerosols present in the cover gas. Thus the knowledge of the amount of aerosol release to the cover gas is important for the evaluation of the source term. The maximum credible cover gas instantaneous source terms deduced from the experiments have led to conservative source terms to be taken into account in safety analysis. Nevertheless modelling attempts of the observed (in-pile or out-of-pile) physico-chemical phenomena have been undertaken for extrapolation to the reactor case. The main topics of this theoretical research are as follows: fission products evaporation in the cover gas (Fz

  8. Optimization of up-flow anaerobic sludge blanket reactor for ...

    African Journals Online (AJOL)

    Optimization of up-flow anaerobic sludge blanket reactor for treatment of composite ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... Granules grown in the bottom part of UASB reactor were more compact and tense ...

  9. Electromagnetic effects involving a tokamak reactor first wall and blanket

    International Nuclear Information System (INIS)

    Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

    1980-01-01

    Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

  10. Blanket options for high-efficiency fusion power

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  11. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  12. Fusion blanket for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Taussig, R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by Ar) utilizing Li 2 O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  13. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1981-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 deg C) of conventional structural materials such as stainless steels. In this project 'two-zone' blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 deg C leading to an overall efficiency estimate of 55 to 60% for this reference case. (author)

  14. Fusion-reactor blanket-material safety-compatibility studies

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

  15. Nuclear characteristics of D-D fusion reactor blankets, (1)

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao; Seki, Yasushi.

    1977-01-01

    Fusion reactors operating on the deuterium (D-D) cycle are considered promising for their freedom from tritium breeding in the blanket. In this paper, neutronic and photonic calculations are undertaken covering several blanket models of the D-D fusion reactor, using presently available data, with a view to comparing the nuclear characteristics of these models, in particular, the nuclear heating rates and their spatial distributions. Nine models are taken up for the study, embodying various combinations of coolant, blanket, structural and reflector materials. About 10 MeV is found to be a typical value for the total nuclear energy deposition per source neutron in the models considered here. The realization of high energy gain is contingent upon finding a favorable combination of blanket composition and configuration. The resulting implications on the thermal design aspect are briefly discussed. (auth.)

  16. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1987-01-01

    A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

  17. Blast venting through blanket material in the HYLIFE ICF reactor

    International Nuclear Information System (INIS)

    Liu, J.C.; Peterson, P.F.; Schrock, V.E.

    1992-01-01

    This work presents a numerical study of blast venting through various blanket configurations in the HYLIFE ICF reactor design. The study uses TSUNAMI -- a multi-dimensional, high-resolution, shock capturing code -- to predict the momentum exchange and gas dynamics for blast venting in complex geometries. In addition, the study presents conservative predictions of wall loading by gas shock and impulse delivered to the protective liquid blanket. Configurations used in the study include both 2700 MJ and 350 MJ fusion yields per pulse for 5 meter and 3 meter radius reactor chambers. For the former, an annular jet array is used for the blanket geometry, while in the latter, both annular jet array as well as slab geometries are used. Results of the study indicate that blast venting and wall loading may be manageable in the HYLIFE-II design by a judicious choice of blanket configuration

  18. Imploding-liner reactor nucleonic studies: the LINUS blanket

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1977-09-01

    Scoping nucleonic studies have been performed for a small imploding-liner fusion reactor concept. Tritium breeding ratio and time-dependent energy deposition rates were the primary parameters of interest in the study. Alloys of Pb and LiPb were considered for the liquid liner (blanket), and tritium breeding was found to be more than adequate with blankets less than 1 m thick. However, neutron leakages into the solid cylinder block surrounding the liquid liner are generally quite high, so considerable effort was concentrated on minimizing these values. Time-dependent calculations reveal that 89% of the energy is deposited in the blanket within 2 μs. Thus, LINUS's blanket should remain intact for the requisite neutron and gamma-ray lifetimes

  19. Application of vanadium alloys to a fusion reactor blanket

    Energy Technology Data Exchange (ETDEWEB)

    Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

    1984-05-01

    Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

  20. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    International Nuclear Information System (INIS)

    Tanigawa, H.; Hirose, T.; Shiba, K.; Kasada, R.; Wakai, E.; Serizawa, H.; Kawahito, Y.; Jitsukawa, S.; Kimura, A.; Kohno, Y.; Kohyama, A.; Katayama, S.; Mori, H.; Nishimoto, K.; Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J.

    2008-01-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed

  1. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

  2. Recent developments in fusion first wall, blanket, and shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-01-01

    This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

  3. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

  4. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-02-01

    A review is given of the features and potentialities of cold-blanket systems, with respect to plasma equilibrium, stability, and reactor technology. The treatment is concentrated on quasi-steady magnetized plasmas confined at moderately high beta values. The cold-blanket concept has specific potentialities as a fusion reactor, e.g. in connection with the desired densities and dimensions of full-scale systems, refuelling, as well as ash and impurity removal, and stability. (author)

  5. Feasibility study of fusion breeding blanket concept employing graphite reflector

    International Nuclear Information System (INIS)

    Cho, Seungyon; Ahn, Mu-Young; Lee, Cheol Woo; Kim, Eung Seon; Park, Yi-Hyun; Lee, Youngmin; Lee, Dong Won

    2015-01-01

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  6. Feasibility study of fusion breeding blanket concept employing graphite reflector

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seungyon, E-mail: sycho@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Woo; Kim, Eung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  7. Overview of first wall/blanket/shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-04-01

    This brief overview of first wall, blanket, and shield technology focuses first on changes and trends in important design issues from the 1970's to the 1980's, then on current perceptions of critical issues in first wall, blanket, and shield design and related technology. The emphasis is on base technology rather than either systems engineering or materials development, on the two primary confinement systems, tokamaks and mirrors, and on production of electricity as the primary goal for development

  8. Applications of the Aqueous Self-Cooled Blanket concept

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.J.; Varsamis, G.; Wrisley, K.; Deutch, L.; Gierszewski, P.

    1986-01-01

    In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids

  9. Progress on DEMO blanket attachment concept with keys and pins

    International Nuclear Information System (INIS)

    Vizvary, Zsolt; Iglesias, Daniel; Cooper, David; Crowe, Robert; Riccardo, Valeria

    2015-01-01

    Highlights: • DEMO blanket attachment system with keys and pins (without using bolts). • Blanket segments are preloaded by progressively designed springs. • Blanket back plate flexibility has a major impact on spring design. • Mechanical analysis of other components indicates no unresolvable issues. • Thermal analysis indicates acceptable temperatures for the support system. - Abstract: The blanket attachment has to cope with gravity, thermal and electromagnetic loads, also it has to be installed and serviced by remote handling. Pre-stressed components suffer from stress relaxation in irradiated environments such as DEMO. To circumvent this problem pre-stressed component should be either avoided or shielded, and where possible keys and pins should be used. This strategy has been proposed for the DEMO multi-module segments (MMS). The blanket segments are held by two tapered keys each, designed to allow thermal expansions while providing contact with the vacuum vessel and to resist the poloidal and radial moments the latter being dominant at 9.1 MNm inboard and 15 MNm outboard. On the top of the blanket segment there is a pin which provides vertical support. At the bottom another vertical support has to lock them in position after installation and manage the pre-load on the segments. The pre-load is required to deal with the electromagnetic loads during disruption. This is provided by a set of springs, which require shielding as they are preloaded. These are sized to cope with the force (3 MN inboard, 1.4 MN outboard) due to halo currents and the toroidal moment which can reverse. Calculations show that the flexibility of the blanket segment itself plays a significant role in defining the required support system. The blanket segment acts as a preloaded spring and it has to be part of the attachment design as well.

  10. Electrical connectors for blanket modules in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I., E-mail: poddubnyyii@nikiet.ru [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Khomiakov, S.; Kolganov, V. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Sadakov, S.; Calcagno, B.; Chappuis, Ph.; Roccella, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Danilov, I.; Leshukov, A.; Strebkov, Y. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Ulrickson, M. [Sandia National Laboratories MS-1129, PO Box 5800, Albuquerque, NM 87185 (United States)

    2014-10-15

    Highlights: • Analysis of static and cyclic strength for L-shaped and Z-shaped ES has been performed. • Analysis results do show that for L-shaped ES static and cyclic strength criteria are not satisfied. • Static and cyclic strength criteria are met well by ES with Z-shaped elastic elements. • ES with Z-shaped elastic elements has been adopted as a new baseline design for ITER. - Abstract: Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak. The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was

  11. Finite element program ARKAS: verification for IAEA benchmark problem analysis on core-wide mechanical analysis of LMFBR cores

    International Nuclear Information System (INIS)

    Nakagawa, M.; Tsuboi, Y.

    1990-01-01

    ''ARKAS'' code verification, with the problems set in the International Working Group on Fast Reactors (IWGFR) Coordinated Research Programme (CRP) on the inter-comparison between liquid metal cooled fast breeder reactor (LMFBR) Core Mechanics Codes, is discussed. The CRP was co-ordinated by the IWGFR around problems set by Dr. R.G. Anderson (UKAEA) and arose from the IWGFR specialists' meeting on The Predictions and Experience of Core Distortion Behaviour (ref. 2). The problems for the verification (''code against code'') and validation (''code against experiment'') were set and calculated by eleven core mechanics codes from nine countries. All the problems have been completed and were solved with the core structural mechanics code ARKAS. Predictions by ARKAS agreed very well with other solutions for the well-defined verification problems. For the validation problems based on Japanese ex-reactor 2-D thermo-elastic experiments, the agreements between measured and calculated values were fairly good. This paper briefly describes the numerical model of the ARKAS code, and discusses some typical results. (author)

  12. Stress analysis of blanket vessel for JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

    A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

  13. Studies on steps affecting tritium residence time in solid blanket

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

    For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

  14. Analyses of Hubble Space Telescope Aluminized-Teflon Multilayer Insulation Blankets Retrieved After 19 Years of Space Exposure

    Science.gov (United States)

    de Groh, Kim K.; Perry, Bruce A.; Mohammed, Jelila S.; Banks, Bruce

    2015-01-01

    Since its launch in April 1990, the Hubble Space Telescope (HST) has made many important observations from its vantage point in low Earth orbit (LEO). However, as seen during five servicing missions, the outer layer of multilayer insulation (MLI) has become increasingly embrittled and has cracked in many areas. In May 2009, during the 5th servicing mission (called SM4), two MLI blankets were replaced with new insulation and the space-exposed MLI blankets were retrieved for degradation analyses by teams at NASA Glenn Research Center (GRC) and NASA Goddard Space Flight Center (GSFC). The retrieved MLI blankets were from Equipment Bay 8, which received direct sunlight, and Equipment Bay 5, which received grazing sunlight. Each blanket was divided into several regions based on environmental exposure and/or physical appearance. The aluminized-Teflon (DuPont, Wilmington, DE) fluorinated ethylene propylene (Al-FEP) outer layers of the retrieved MLI blankets have been analyzed for changes in optical, physical, and mechanical properties, along with chemical and morphological changes. Pristine and as-retrieved samples (materials) were heat treated to help understand degradation mechanisms. When compared to pristine material, the analyses have shown how the Al-FEP was severely affected by the space environment. Most notably, the Al-FEP was highly embrittled, fracturing like glass at strains of 1 to 8 percent. Across all measured properties, more significant degradation was observed for Bay 8 material as compared to Bay 5 material. This paper reviews the tensile and bend-test properties, density, thickness, solar absorptance, thermal emittance, x-ray photoelectron spectroscopy (XPS) and energy dispersive spectroscopy (EDS) elemental composition measurements, surface and crack morphologies, and atomic oxygen erosion yields of the Al-FEP outer layer of the retrieved HST blankets after 19 years of space exposure.

  15. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

    2011-01-01

    Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

  16. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  17. Conceptual design of solid breeder blanket system cooled by supercritical water

    International Nuclear Information System (INIS)

    Enoeda, Mikio; Akiba, Masato; Ohara, Yoshihiro

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li 2 TiO 3 or Li 2 O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  18. In plain sight: the Chesapeake Bay crater ejecta blanket

    Science.gov (United States)

    Griscom, D. L.

    2012-02-01

    idealized calculation of the CBIS ejecta-blanket elevation profile minutes after the impact was carried out founded on well established rules for explosion and impact-generated craters. This profile is shown here to match the volume of the upland deposits ≥170 km from the crater center. Closer to the crater, much of the "postdicted" ejecta blanket has clearly been removed by erosion. Nevertheless the Shirley and fossil-free Bacons Castle Formations, located between the upland deposits and the CBIS interior and veneering the present day surface with units ∼10-20 m deep, are respectively identified as curtain- and excavation-phase ejecta. The neritic-fossil-bearing Calvert Formation external to the crater is deduced to be of Eocene age (as opposed to early Miocene as currently believed), preserved by the armoring effects of the overlying CBIS ejecta composed of the (distal) upland deposits and the (proximal) Bacons Castle Formation. The lithofacies of the in-crater Calvert Formation can only have resulted from inward mass wasting of the postdicted ejecta blanket, vestiges of which (i.e. the Bacons Castle and Shirley Formations) still overlap the crater rim and sag into its interior, consistent with this expectation. Because there appear to be a total of ∼10 000 km2 of CBIS ejecta lying on the present-day surface, future research should center the stratigraphic, lithologic, and petrologic properties of these ejecta versus both radial distance from the crater center (to identify ejecta from different ejection stages) and circumferentially at fixed radial distances (to detect possible anisotropies relating the impact angle and direction of approach of the impactor). The geological units described here may comprise the best preserved, and certainly the most accessible, ejecta blanket of a major crater on the Earth's surface and therefore promise to be a boon to the field of impact geology. As a corollary, a major revision of the current stratigraphic column of the M

  19. The RCC-MR design code for LMFBR components. A useful basic for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1985-11-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials (Stainless steels), temperature service level (550-600 0 C), loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain

  20. The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1986-01-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)