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Sample records for high burn-up fbr

  1. Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

    2005-01-01

    The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

  2. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  3. Ultrasonic measurement of high burn-up fuel elastic properties

    International Nuclear Information System (INIS)

    Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.

    2006-01-01

    The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment

  4. Test of calorimetry for high burn-up plutonium

    International Nuclear Information System (INIS)

    Beets, C.; Carchon, R.; Fettweis, P.

    1984-01-01

    In recent times, the interest of applying calorimetry for safeguards purpose is steadily increasing. Calorimetric measurements have been performed on a set of high burn-up (25000 MWd/t) Pu samples, ranging in mass between 60 g and 2.5 kg Pu, distributed as PuO 2 powder embedded in stainless steel containers. The powers produced by these containers ranged between 0.8 W and 36 W. The calorimeter used was the Mound 150 type, and the isotopics and the Am content have been determined earlier by mass spectroscopy, completed with α and γ counting, and were later verified by the same methods. Watts/gram measurements were made on twelve 60 g samples of the same plutonium lot to demonstrate the Pu elemental and isotopic homogeneity, and hence, its suitability for subsequent NDA experiments. These samples were also measured in a stacked way to fill up the mass and wattage gaps between 60 g (0.8W) and 1 kg (14W). Calorimetric assay values, obtained with both isotopic measurements are discussed

  5. Review of high burn-up RIA and LOCA database and criteria

    International Nuclear Information System (INIS)

    Vitanza, C.; Hrehor, M.

    2006-01-01

    This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

  6. Mesoscopic approach to describe high burn-up fuel behaviour

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1999-01-01

    The grain sub-division and the rim structure formation are new phenomena for LWR fuel engineering. The consequence of these are now under investigation in several international programs such as HBRP (High Burnup Rim Project) of CRIEPI, NFIR of EPRI, and EdF/CEA program in France. The theoretical understanding of this phenomenon is underway. Here, the process is peculiar in the following points; (1) majority of the domain of the material are changed to a new morphology after the restructuring, (2) the final size of the new grains is around 0.1 μm which is neither atomic scale nor macroscopic scale. (3) the morphology of the restructured domain indicates fractal like feature which indicates complex process is under-taken. From the first feature, the process is similar to phase transitions or metallographic transformations. However, as the crystallographic structure has no change before and after the restructuring, it is not the phase transition nor the transformation of atomic scale instability. The focus could be put on the material transport of mesoscopic scale which create the peculiar morphology. Indeed there are flows of energy and disturbances in crystallographic structure in nuclear materials on duty. Although the fission energy is 10 4 larger than the formation energy of the defects, thanks to the stability of the selected material, most of energy is thermalized without crystallographic instability. Little remained energy creates flows of disturbances and the new structure is a consequence of ordering process driven by these flows of disturbances. Therefore this phenomenon is a good example to study cooperative ordering process in physics of materials. This paper presents some of present understandings of the rim structure formation based on the mesoscopic mechanistic theories. Possible future development is also proposed (author) (ml)

  7. High-burn-up fuels for fast reactors. Past experience and novel applications

    International Nuclear Information System (INIS)

    Weaver, Kevan D.; Gilleland, John; Whitmer, Charles; Zimmerman, George

    2009-01-01

    Fast reactors in the U.S. routinely achieved fuel burn-ups of 10%, with some fuel able to reach peak burn-ups of 20%, notably in the Experimental Breeder Reactor II and the Fast Flux Test Facility. Maximum burn-up has historically been constrained by chemical and mechanical interactions between the fuel and its cladding, and to some extent by radiation damage and thermal effects (e.g., radiation-induced creep, thermal creep, and radiation embrittlement) that cause the cladding to weaken. Although fast reactors have used several kinds of fuel - including oxide, metal alloy, carbide, and nitride - the vast majority of experience with fast reactors has been using oxide (including mixed oxide) and metal-alloy fuels based on uranium. Our understanding of high-burn-up operation is also limited by the fact that breeder reactor programs have historically assumed that their fuel would eventually undergo reprocessing; the programs thus have not made high burn-up a top priority. Recently a set of novel designs have emerged for fast reactors that require little initial enrichment and no reprocessing. These reactors exploit a concept known as a traveling wave (sometimes referred to as a breed-and-burn wave, fission wave, or nuclear-burning wave). By breeding and using its own fuel in place as it operates, a traveling-wave reactor can obtain burn-ups that approach 50%, well beyond the current base of knowledge and experience. Our computational work on the physics of traveling-wave reactors shows that they require metal-alloy fuel to provide the margins of reactivity necessary to sustain a breed-and-burn wave. This paper reviews operating experience with high-burn-up fuels and the technical feasibility of moving to a qualitatively new burn-up regime. We discuss our calculations on traveling-wave reactors, including those concerning the possible use of thorium. The challenges associated with high burn-up and fluence in fuels and materials are also discussed. (author)

  8. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    International Nuclear Information System (INIS)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

    2003-01-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

  9. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

    2003-07-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

  10. Burn-Up Determination by High Resolution Gamma Spectrometry: Fission Product Migration Studies

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Blackadder, W H; Ronqvist, N

    1967-04-15

    The migration of solid fission products, in particular caesium and ruthenium, in high temperature oxide fuel can create a severe problem during the application of non-destructive burn-up methods employing gamma spectrometry, since caesium-137 is otherwise the most convenient long-lived burn-up monitor and ruthenium-106 can be used to distinguish between fissions in U-235 and Pu-239. As part of an experimental programme to develop burn-up methods, gamma scanning experiments have been performed on slices of irradiated UO{sub 2} pellets using a lithium-drifted germanium detector. The usefulness of the technique for migration studies has been demonstrated by comparing the fission product distribution curves across the specimen diameters with the microstructure of the specimens after polishing and etching.

  11. Development of high performance liquid chromatography for rapid determination of burn-up of nuclear fuels

    International Nuclear Information System (INIS)

    Joseph, M.; Karunasagar, D.; Saha, B.

    1996-01-01

    Burn-up an important parameter during evaluation of the performance of any nuclear fuel. Among the various techniques available, the preferred one for its determination is based on accurate measurement of a suitable fission product monitor and the residual heavy elements. Since isotopes of rare earth elements are generally used as burn-up monitors, conditions were standardized for rapid separation (within 15 minutes) of light rare earths using high performance liquid chromatography based on either anion exchange (Partisil 10 SAX) in methanol-nitric acid medium or by cation exchange on a reverse phase column (Spherisorb 5-ODS-2 or Supelcosil LC-18) dynamically modified with 1-octane sulfonate or camphor-10-sulfonic acid (β). Both these methods were assessed for separation of individual fission product rare earths from their mixtures. A new approach has been examined in detail for rapid assay of neodymium, which appears promising for faster and accurate measurement of burn-up. (author)

  12. Fission gas release from UO2 pellet fuel at high burn-up

    International Nuclear Information System (INIS)

    Vitanza, C.; Kolstad, E.; Graziani, U.

    1979-01-01

    Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

  13. Study on the thermal-hydraulic stability of high burn up STEP III fuel in Japan

    International Nuclear Information System (INIS)

    Ishikawa, M.; Kitamura, H.; Toba, A.; Omoto, A.

    2004-01-01

    Japanese BWR utilities have performed a joint study of the Thermal Hydraulic Stability of High Burn up STEP III Fuel. In this study, the parametric dependency of thermal hydraulic stability threshold was obtained. It was confirmed through experiments that the STEP III Fuel has sufficient stability characteristics. (author)

  14. Effect of high burn-up and MOX fuel on reprocessing, vitrification and disposal of PWR and BWR spent fuels based on accurate burn-up calculation

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, T.; Iwasaki, T.; Wada, K. [Tohoku Univ., Graduate School of Engineering, Dept. of Quantum Science and Energy Engineering, Sendai 980-8579 (Japan); Suyama, K. [Japan Atomic Energy Agency, Shirakata-Shirane 2-4, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2006-07-01

    To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)

  15. High burn-up structure in nuclear fuel: impact on fuel behavior - 4005

    International Nuclear Information System (INIS)

    Noirot, J.; Pontillon, Y.; Zacharie-Aubrun, I.; Hanifi, K.; Bienvenu, P.; Lamontagne, J.; Desgranges, L.

    2016-01-01

    When UO 2 and (U,Pu)O 2 fuels locally reach high burn-up, a major change in the microstructure takes place. The initial grains are replaced by thousands of much smaller grains, fission gases form micrometric bubbles and metallic fission products form precipitates. This occurs typically at the rim of the pellets and in heterogeneous MOX fuel Pu rich agglomerates. The high burn-up at the rim of the pellets is due to a high capture of epithermal neutrons by 238 U leading locally to a higher concentration of fissile Pu than in the rest of the pellet. In the heterogeneous MOX fuels, this rim effect is also active, but most of the high burn-up structure (HBS) formation is linked to the high local concentration of fissile Pu in the Pu agglomerates. This Pu distribution leads to sharp borders between HBS and non-HBS areas. It has been shown that the size of the new grains, of the bubbles and of the precipitates increase with the irradiation local temperatures. Other parameters have been shown to have an influence on the HBS initiation threshold, such as the irradiation density rate, the fuel composition with an effect of the Pu presence, but also of the Gd concentration in poisoned fuels, some of the studied additives, like Cr, and, maybe some of the impurities. It has been shown by indirect and direct approaches that HBS formation is not the main contributor to the increase of fission gas release at high burn-up and that the HBS areas are not the main source of the released gases. The impact of HBS on the fuel behavior during ramp on high burn-up fuels is still unclear. This short paper is followed by the slides of the presentation

  16. Cellular automata approach to investigation of high burn-up structures in nuclear reactor fuel

    International Nuclear Information System (INIS)

    Akishina, E.P.; Ivanov, V.V.; Kostenko, B.F.

    2005-01-01

    Micrographs of uranium dioxide (UO 2 ) corresponding to exposure times in reactor during 323, 953, 971, 1266 and 1642 full power days were investigated. The micrographs were converted into digital files isomorphous to cellular automata (CA) checkerboards. Such a representation of the fuel structure provides efficient tools for its dynamics simulation in terms of primary 'entities' imprinted in the micrographs. Besides, it also ensures a possibility of very effective micrograph processing by CA means. Interconnection between the description of fuel burn-up development and some exactly soluble models is ascertained. Evidences for existence of self-organization in the fuel at high burn-ups were established. The fractal dimension of microstructures is found to be an important characteristic describing the degree of radiation destructions

  17. Fission gas release at high burn-up: beyond the standard diffusion model

    International Nuclear Information System (INIS)

    Landskron, H.; Sontheimer, F.; Billaux, M.R.

    2002-01-01

    At high burn-up standard diffusion models describing the release of fission gases from nuclear fuel must be extended to describe the experimental loss of xenon observed in the fuel matrix of the rim zone. Marked improvements of the prediction of integral fission gas release of fuel rods as well as of radial fission gas profiles in fuel pellets are achieved by using a saturation concept to describe fission gas behaviour not only in the pellet rim but also as an additional fission gas path in the whole pellet. (author)

  18. EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

    International Nuclear Information System (INIS)

    Teague, Melissa C; Gorman, Brian P.; Miller, Brandon D; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel

  19. An investigation into fuel pulverization with specific reference to high burn-up LOCA

    International Nuclear Information System (INIS)

    Yagnik, Suresh; Turnbull, James; Noirot, Jean; Walker, Clive; Hallstadius, Lars; Waeckel, N.; Blanpain, P.

    2014-01-01

    To investigate the phenomenon of high burn-up fuel pellet material potentially disintegrating into powder under a rapid temperature transient, such as in a LOCA-type accident scenario, two independent scoping studies were commissioned. The first was to investigate the effect of hydrostatic restraint pressure on Fission Gas Release (FGR) from small samples of highly irradiated fuel (71 MWd/kgU) during a series of rapid temperature ramps. Experimentally, when the FGR increased rapidly during the temperature transients, the fuel was assumed to be 'pulverized', i.e., fragmented into powder. In the second series of experiments, laser heating of small samples was used to investigate the temperature at which fuel pulverization was initiated. Subsequent to fuel disintegration, there was always a spectrum of particle sizes present. The significance of this observation was recognized in the context of extended burn-up operation in commercial reactors. Based on the observation from these investigations, a fuel fragmentation threshold has been discussed and developed. We conclude that fuel disintegration could be of potential importance in limiting the performance and productive lifetime of nuclear fuel. However, since only fuel closely adjacent to ballooned or ruptured cladding would be released in a LOCA-type transient, expulsion of pulverized fuel from the ruptured fuel rod is not considered a safety issue; cooling of the defected assembly remains possible and there is no issue with respect to local criticality. (author)

  20. IFPE/HBEP REV.1, Battelle's High Burn-Up Effects Programme for Fuel Performance

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2002-01-01

    Description: It contains data from phase 2 and 3 on fabrication, dimensions, fuel and cladding properties and composition, reactor conditions and Post Irradiation Examination (PIE) data of the High Burn-up Effects Programme (HBEP) carried out at the Battelle North-west Laboratories. Each data set contains a full irradiation history with clad temperature and local power listed for each rod at 5, 10 or 12 axial zones as a function of cumulative time to the end of the given time interval over which the power has been constant. Data is provided for 45 rods from phase 2 and 36 rods from phase 3. The different rods have been manufactured by: ASEA/TVO, BN, BNFL, FBFC, FRA/CEA, GE, KWU/CE, WEC

  1. Burn-up measurements on nuclear reactor fuels using high performance liquid chromatography

    International Nuclear Information System (INIS)

    Sivaraman, N.; Subramaniam, S.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2002-01-01

    Burn-up measurements on thermal as well as fast reactor fuels were carried out using high performance liquid chromatography (HPLC). A column chromatographic technique using di-(2-ethylhexyl) phosphoric acid (HDEHP) coated column was employed for the isolation of lanthanides from uranium, plutonium and other fission products. Ion-pair HPLC was used for the separation of individual lanthanides. The atom percent fissions were calculated from the concentrations of the lanthanide (neodymium in the case of thermal reactor and lanthanum for the fast reactor fuels) and from uranium and plutonium contents of the dissolver solutions. The HPLC method was also used for determining the fractional fissions from uranium and plutonium for the thermal reactor fuel. (author)

  2. Micrographic study on distribution of fission products in high burn-up metallic alloy fuel

    International Nuclear Information System (INIS)

    Kolay, S.; Basu, M.; Das, D.

    2012-01-01

    One of the important mandates in the three-stage nuclear power generation programme of India is to utilize uranium-plutonium based alloy fuels in enabling shorter doubling time for breeding of the fissile isotopes ( 239 Pu and 233 U ) to be used in thorium based driver fuel in the third stage. Reported information shows the successful performance of alloy fuel with somewhat porous matrix in achieving 10-15 atom% burnup. The porosity and microstructure of these alloys are strongly dependent on their composition and phases present. Porosity also influences the extent of fuel swelling and gas release. So to assess fuel performance and fuel integrity under high burn-up condition it is essential to have knowledge about the new phases formed and their redistribution that occurs as a result of inter-diffusion and temperature gradient. This study addresses these issues taking the base alloy U-10 wt %Zr

  3. Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

    International Nuclear Information System (INIS)

    Mogensen, M.; Walker, C.T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU. (orig.)

  4. Experimental modeling of high burn-up structure in SIMFUEL with ion irradiation

    International Nuclear Information System (INIS)

    Baranov, V.; Isaenkova, M.; Lunev, A.; Tenishev, A.; Khlunov, A.

    2013-01-01

    Experiments are conducted to simulate high burn-up structure in accelerator conditions. Three ion irradiation schemes are used: 1. Xe 27+ 160 MeV up to 5x10 15 cm -2 (thermal spikes). 2. Xe 16+ 320 keV up to 1x10 17 cm -2 (collision cascades). 3. He + 20 keV up to 5,5x10 17 cm -2 (implantation stage). Structural characterization performed by scanning electron microscopy, X-ray analysis and atomic force microscopy revealed prominent grain refinement in case of Xe 27+ irradiation. Artificial energy variation for incident ions showed varying size of subgrains. At maximum energy of incident ions, subgrain size amounts ∼ 320 nm. Moving to the edge of irradiated region changes the size to ∼ 170 nm. Typical size of coherent scattering regions matches subgrain size for high-energy irradiation. Low-energy irradiation results in less significant structural changes: flaky structure at random sites for samples irradiated with low-energy xenon ions and bubble nucleation for helium irradiation. Dislocation density increases significantly, and it is shown that a single fluence dependence exists for low- and high-energy irradiation. (authors)

  5. Simulated LOCA Test and Characterization Study Related to High Burn-Up Issue

    International Nuclear Information System (INIS)

    Park, D. J.; Jung, Y. I.; Choi, B. K.; Park, S. Y.; Kim, H. G.; Park, J. Y.

    2012-01-01

    For the safety evaluation of fuel cladding during the injection of emergency core coolant, simulated Loss-of-coolant accident (LOCA) test was performed by using Zircaloy-4 fuel cladding samples. Zircaloy-4 tube samples with and without prehydring were oxidized in a steam environment with the test temperature of 1200 .deg. C. Prehydrided cladding was prepared from as-fabricated Zircaloy-4 to study the effects of hydrogen on mechanical properties of cladding during high temperature oxidation and quench conditions. In order to measure the ductility of the tube samples embrittled by quenching water, ring compression test was carried out by using 8 mm ring sample sectioned from oxidized tube sample and microstructural analysis was also performed after simulated LOCA test. The results showed that hydrogen increases oxygen solubility and pickup rate in the beta layer. This reduces ductility of prehydrided fuel cladding compared with as-fabricated cladding. Trend in ductility decrease for prehydrided sample under simulated LOCA condition was very similar with data obtained from tests conducted using irradiated high burn-up fuel claddings

  6. TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

    2014-04-01

    As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

  7. Non-instrumented capsule design of HANARO irradiation test for the high burn-up large grain UO2 pellets

    International Nuclear Information System (INIS)

    Kim, D. H.; Lee, C. B.; Oh, D. S.

    2001-01-01

    Non-instrumented capsule was designed to irradiate the large grain UO 2 pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. UO 2 pelletes will be irradiated up to the burn-up higher than 70 MWD/kgU in HANARO. To irradiate the UO 2 pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. In addition, to satisfy the safety criteria of HANARO such as prevention of ONB(Onset of Nucleate Boiling), fuel melting and wear damage of the capsule during the long term irradiation, design of the non-instrumented capsule was optimized

  8. Total surface area change of Uranium dioxide fuel in function of burn-up and its impact on fission gas release during neutron irradiation for small, intermediate and high burn-up

    International Nuclear Information System (INIS)

    Szuta, M.

    2011-01-01

    In the early published papers it was observed that the fractional fission gas release from the specimen have a tendency to increase with the total surface area of the specimen - a fairy linear relationship was indicated. Moreover it was observed that the increase of total surface area during irradiation occurs in the result of connection the closed porosity with the open porosity what in turn causes the increase of fission gas release. These observations let us surmise that the process of knock-out release is the most significant process of fission gas release since its quantity is proportional to the total surface area. Review of the experiments related to the increase of total surface area in function of burn-up is presented in the paper. For very high burn-up the process of grain sub-division (polygonization) occurs under condition that the temperature of irradiated fuel lies below the temperature of grain re-crystallization. Simultaneously with the process of polygonization, the increase in local porosity and the decrease in local density in function of burn-up occurs, which leads to the increase of total surface area. It is suggested that the same processes take place in the transformed fuel as in the original fuel, with the difference that the total surface area is so big that the whole fuel can be treated as that affected by the knock-out process. This leads to explanation of the experimental data that for very high burn-up (>120 MWd/kgU) the concentration of xenon is constant. An explanation of the grain subdivision process in function of burn-up in the 'athermal' rim region in terms of total surface area, initial grain size and knock-out release is undertaken. Correlation of the threshold burn-up, the local fission gas concentration, local total surface area, initial and local grain size and burn-up in the rim region is expected. (author)

  9. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...

  10. UO2 fuel behaviour at rod burn-ups up to 105 MWd/kgHM. A review of 10 years of high burn-up examinations commissioned by AREVA NP

    International Nuclear Information System (INIS)

    Goll, W.; Hoffmann, P.B.; Hellwig, C.; Sauser, W.; Spino, J.; Walker, C.T.

    2007-01-01

    Irradiation experience gained on fuel rods with burn-ups greater than 60 MWd/kgHM irradiated in the Nuclear Power Plant Goesgen, Switzerland, is described. Emphasis is placed on the fuel behaviour, which has been analysed by hot cell examinations at the Institute for Transuranium Elements and the Paul-Scherrer-Institute. Above 60 MWd/kgHM, the so-called high burn-up structure (HBS) forms and the fission gas release increases with burn-up and rod power. Examinations performed in the outer region of the fuel revealed that most if not all of the fission gas created was retained in the HBS, even at 25% porosity. Furthermore, the HBS has a relatively low swelling rate, greatly increased plasticity, and its thermal conductivity is higher than expected from the porosity. The post-irradiation examinations showed that the HBS has no detrimental effects on the performance of stationary irradiated PWR fuel irradiated to the high burn-ups that can be achieved with 5 wt% U-235 enrichment. On the contrary, the HBS results in fuel performance that is generally better than it would have been if the HBS had not formed. (orig.)

  11. The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

    2002-01-01

    In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

  12. Modelling of pore coarsening in the high burn-up structure of UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S.; Tarasov, V.I., E-mail: tarasov@ibrae.ac.ru

    2017-05-15

    The model for coalescence of randomly distributed immobile pores owing to their growth and impingement, applied by the authors earlier to consideration of the porosity evolution in the high burn-up structure (HBS) at the UO{sub 2} fuel pellet periphery (rim zone), was further developed and validated. Predictions of the original model, taking into consideration only binary impingements of growing immobile pores, qualitatively correctly describe the decrease of the pore number density with the increase of the fractional porosity, however notably underestimate the coalescence rate at high burn-ups attained in the outmost region of the rim zone. In order to overcome this discrepancy, the next approximation of the model taking into consideration triple impingements of growing pores was developed. The advanced model provides a reasonable consent with experimental data, thus demonstrating the validity of the proposed pore coarsening mechanism in the HBS.

  13. LWR high burn-up operation and MOX introduction. Fuel cycle performance from the viewpoint of waste management

    International Nuclear Information System (INIS)

    Inagaki, Yaohiro; Iwasaki, Tomohiko; Niibori, Yuichi; Sato, Seichi; Ohe, Toshiaki; Kato, Kazuyuki; Torikai, Seishi; Nagasaki, Shinya; Kitayama, Kazumi

    2009-01-01

    From the viewpoint of waste management, a quantitative evaluation of LWR nuclear fuel cycle system performance was carried out, considering both higher burn-up operation of UO 2 fuel coupled with the introduction of MOX fuel. A major parameter to quantify this performance is the number of high-level waste (HLW) glass units generated per GWd (gigawatt-day based on reactor thermal power generation before electrical conversion). This parameter was evaluated for each system up to a maximum burn-up of 70GWd/THM (gigawatt-day per ton of heavy metal) assuming current conventional reprocessing and vitrification conditions where the waste loading of glass is restricted by the heat generation rate, the MoO 3 content, or the noble metal content. The results showed that higher burn-up operation has no significant influence on the number of glass units generated per GWd for UO 2 fuel, though the number of glass units per THM increases linearly with burn-up and is restricted by the heat generation rate. On the other hand, the introduction of MOX fuel causes the number of glass units per GWd to double owing to the increase in the heat generation rate. An extended cooling period of the spent fuel prior to reprocessing effectively reduces the heat generation rate for UO 2 fuel, while a separation of minor actinides (Np, Am, and Cm) from the high-level waste provides additional reduction for MOX fuel. However, neither of these leads to a substantial reduction in the number of glass units, since the MoO 3 content or the noble metal content restricts the number of glass units rather than the heat generation rate. These results suggest that both the MoO 3 content and the noble metal content provide the key to reducing the amount of waste glass that is generated, leading to an overall improvement in fuel cycle system performance. (author)

  14. On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

    International Nuclear Information System (INIS)

    Walker, C.T.; Staicu, D.; Sheindlin, M.; Papaioannou, D.; Goll, W.; Sontheimer, F.

    2006-01-01

    A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation

  15. Burn-Up Determination by High Resolution Gamma Spectrometry: Axial and Diametral Scanning Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Blackadder, W H; Ronqvist, N

    1967-02-15

    In the gamma spectrometric determination of burn-up the use of a single fission product as a monitor of the specimen fission rate is subject to errors caused by activity saturation or, in certain cases, fission product migration. Results are presented of experiments in which all the resolvable gamma peaks in the fission product spectrum have been used to calculate the fission rate; these results form a pattern which reflect errors in the literature values of the gamma branching ratios, fission yields etc., and also represent a series of empirical correction factors. Axial and diametral scanning experiments on a long-irradiated low-enrichment fuel element are also described and demonstrate that it is possible to differentiate between fissions in U-235 and in Pu-239 respectively by means of the ratios of the Ru-106 activity to the activities of the other fission products.

  16. Microstructure Changes in a high burn up Spent Fuel (57,900 MWd/tU)

    International Nuclear Information System (INIS)

    Park, Yang Soon; Kwon, Hyoung Mun; Seo, Hang Seok; Ha, Yeong Keong; Song, Kyuseok

    2009-01-01

    In the nuclear industry, an increase in the burn up and the residence time of fuels is being considered because of the advantages in the fuel cycle cost and the spent fuel management. But, it leads to structural changes in an outer zone (rim) of a UO 2 pellet within a few hundreds of micrometers in thickness. Despite its thin layer, this rim would determine the thermal behavior of a fuel. Therefore, to identify a rim zone effect, the microstructures such as the pores, the grains and the UO 2 lattice size have been investigated by many researchers. In this study, the microstructure changes in the rim of a UO 2 spent fuel, the corrosion layer of a Zry-4 cladding and the interface between a fuel and a cladding were investigated by a micro-XRD and a SEM

  17. A contribution to the understanding of the high burn-up structure formation in nuclear fuels

    International Nuclear Information System (INIS)

    Jonnet, J.

    2007-01-01

    An increase of the discharge burn-up of UO 2 nuclear fuels in the light water reactors results in the appearance of a change of microscopic structure, called HBS. Although well characterised experimentally, important points on the mechanisms of its formation remain to be cleared up. In order to answer these questions, a study of the contribution of the dislocation-type defects was conducted. In a first part, a calculation method of the stress field associated with periodic configurations of dislocations was developed. The method was applied to the cases of edge dislocation pile-up and wall, for which an explicit expression of the internal stress potential was obtained. Through the study of other examples of dislocation configurations, it was highlighted that this method also allows the calculation of any periodic dislocation configuration. In a second part, the evolution of interstitial-type dislocation loops was studied in UO 2 fuel samples doped with 10% in mass of alpha emitters. The experimental loop size distributions were obtained for these samples stored during 4 and 7 years at room temperature. Kinetic equations are proposed in order to study the influence of the resolution process of interstitials from a loop back to the matrix due to an impact with the recoil atom 234 U, as well as the coalescence of two interstitial loops that can diffuse by a volume mechanism. The application of the model shows that the two processes must be considered in the study of the evolution of radiation damage. (author)

  18. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However

  19. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel

    International Nuclear Information System (INIS)

    Horvath, M. I.

    2008-01-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However, relevant Xe

  20. Thermomechanical behavior and modeling of zircaloy cladding tubes from an unirradiated state to high burn-up

    International Nuclear Information System (INIS)

    Schaeffler-Le Pichon, I.; Geyer, P.; Bouffioux, P.

    1997-01-01

    Creep laws are nowadays commonly used to simulate the fuel rod response to the solicitations it faces during its life. These laws are sufficient for describing the base operating conditions (where only creep appears), but they have to be improved for power ramp conditions (where hardening and relaxation appear). The modification due to a neutronic irradiation of the thermomechanical behavior of stress-relieved Zircaloy 4 fuel tubes that have been analysed for five different fluences ranging from a non-irradiated material to a material for which the combustion rate was very high is presented. In the second part, a viscoplastic model able to simulate, for different isotherms, out-of-flux anisotropic mechanical behavior of the cladding tubes irradiated until high burn-up is proposed. Finally, results of numerical simulations show the ability of the model to reproduce the totality of the thermomechanical experiments. (author)

  1. On the rate determining step in fission gas release from high burn-up water reactor fuel during power transients

    International Nuclear Information System (INIS)

    Walker, C.T.; Mogensen, M.

    1987-01-01

    The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved. (orig.)

  2. Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Yuji, E-mail: fukaya.yuji@jaea.go.jp; Nishihara, Tetsuo

    2016-10-15

    Highlights: • We evaluate the number of canisters and its footprint for HTGR. • We proposed new waste loading method for direct disposal of HTGR. • HTGR can significantly reduce HLW volume compared with LWR. - Abstract: Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

  3. SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

    2014-04-01

    During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.

  4. Instant release fraction and matrix release of high burn-up UO{sub 2} spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Energy Technology Data Exchange (ETDEWEB)

    Serrano-Purroy, D., E-mail: Daniel.serrano-purroy@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Clarens, F.; Gonzalez-Robles, E. [CTM Centre Tecnologic, Avda. Bases de Manresa 1, 08240 Barcelona (Spain); Glatz, J.P.; Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Pablo, J. de [CTM Centre Tecnologic, Avda. Bases de Manresa 1, 08240 Barcelona (Spain); Department of Chemical Engineering, Universitat Politecnica de Catalunya, Avda. Diagonal 647, 08028 Barcelona (Spain); Casas, I.; Gimenez, J. [Department of Chemical Engineering, Universitat Politecnica de Catalunya, Avda. Diagonal 647, 08028 Barcelona (Spain); Martinez-Esparza, A. [ENRESA, C/Emilio Vargas 7, 28043 Madrid (Spain)

    2012-08-15

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  5. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Energy Technology Data Exchange (ETDEWEB)

    Lemmens, K., E-mail: klemmens@sckcen.be [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); González-Robles, E.; Kienzler, B. [Karlsruhe Institute of Technology Institute for Nuclear Waste Disposal (KIT-INE), PO Box 3640, D-76021 Karlsruhe (Germany); Curti, E. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Serrano-Purroy, D. [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, PO Box 2340, D-76125 Karlsruhe (Germany); Sureda, R.; Martínez-Torrents, A. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Roth, O. [Studsvik, Nuclear AB, 611 82 Nyköping (Sweden); Slonszki, E. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary); Mennecart, T. [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Günther-Leopold, I. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Hózer, Z. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary)

    2017-02-15

    The instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45–63 GWd/t{sub HM} and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride – bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H{sub 2} atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways. - Highlights: • Leach tests were performed to study the instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel. • In these tests, the fission gas release given by the operator was a pessimistic estimator of the iodine and cesium release. • Iodine and cesium release is proportional to linear power rating beyond 200 W cm{sup −1}. • Closure of the fuel-cladding gap at high burn-up slows down the release. • The release rate decreases following an exponential equation.

  6. Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

    Energy Technology Data Exchange (ETDEWEB)

    Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01

    The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)

  7. Development of high-strength aluminum alloys for basket in transport and storage cask for high burn-up spent fuel

    International Nuclear Information System (INIS)

    Maeguchi, T.; Sakaguchi, Y.; Kamiwaki, Y.; Ishii, M.; Yamamoto, T.

    2004-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has developed high-strength borated aluminum alloys (high-strength B-Al alloys), suitable for application to baskets in transport and storage casks for high burn-up spent fuels. Aluminum is a suitable base material for the baskets due to its low density and high thermal conductivity. The aluminum basket would reduce weight of the cask, and effectively release heat generated by spent fuels. MHI had already developed borated aluminum alloys (high-toughness B-Al alloy), and registered them as ASME Code Case ''N-673''. However, there has been a strong demand for basket materials with higher strength in the case of MSF (Mitsubishi Spent Fuel) casks for high-burn up spent fuels, since the basket is required to stand up to higher stress at higher temperature. The high-strength basket material enables the design of a compact cask under a limitation of total size and weight. MHI has developed novel high-strength B-Al alloys which meet these requirements, based on a new manufacturing process. The outline of mechanical and metallurgical characteristics of the high-strength B-Al alloys is described in this paper

  8. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO2 and MOX spent fuels

    International Nuclear Information System (INIS)

    Sasahara, Akihiro; Matsumura, Tetsuo; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-01-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO 2 spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, a) isotopic analysis, b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  9. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-03-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  10. Peculiarities of highly burned-up NPP SNF reprocessing and new approach to simulation of solvent extraction processes

    Energy Technology Data Exchange (ETDEWEB)

    Fedorov, Y.S.; Zilberman, B.Y.; Goletskiy, N.D.; Puzikov, E.A.; Ryabkov, D.V.; Rodionov, S.A.; Beznosyuk, V.I.; Petrov, Y.Y.; Saprykin, V.F.; Murzin, A.A.; Bibichev, B.A.; Aloy, A.S.; Kudinov, A.S.; Blazheva, I.V. [RPA ' V.G.Khlopin Radium Institute' , 28, 2 Murinsky av., St-Petersburg, 194 021 (Russian Federation); Kurenkov, N.V. [Institute of Industrial Nuclear Technology NRNU MEPHI, 31, Kashirskoye shosse, Moscow, 115409 (Russian Federation)

    2013-07-01

    Substantiation, general description and performance characteristics of a reprocessing flowsheet for WWER-1000 spent fuel with burn-up >60 GW*day/t U is given. Pu and U losses were <0.1%, separation factor > 10{sup 4}; their decontamination factor from γ-emitting fission products was 4*10{sup 4} and 3*10{sup 7}, respectively. Zr, Tc, Np removal was >98% at U and Pu losses <0.05%. A new approach to simulation of extraction equilibrium has been developed. It is based on a set of simultaneous chemical reactions characterized by apparent concentration constants. A software package was created for simulation of spent fuel component distribution in multistage countercurrent extraction processes in the presence of salting out agents. (authors)

  11. Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States); Andrews, Nathan C., E-mail: nandrews@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States)

    2013-05-15

    Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

  12. Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

    International Nuclear Information System (INIS)

    Karahan, Aydın; Andrews, Nathan C.

    2013-01-01

    Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

  13. IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2003-01-01

    Description: The fuel segments for the high burn-up integral rod behaviour test IFA-597 were taken from fuel rod 33-25065, which was irradiated in the Ringhals 1 BWR for approximately 12 years. The irradiation of this rod and its sibling rod 33-25046 was performed in two stages. During the first irradiation, 1980 to 1986, the rods were part of Ringhals assembly 6477 and an approximate rod averaged burn-up of 31 MWd/kg UO 2 was reached. The rods were then placed into fuel assembly 9902 for a second period of irradiation from 1986 to 1992. The location of the fuel rods 33-25065 and 33-25046 in this assembly were in positions 9902/D and 9902/E4 respectively. A final rod averaged burn-up of 52 MWd/kg UO 2 was achieved. The burn-up at the location of the Halden segments was estimated as 59 MWd/kg UO 2 , well beyond the formation of High Burn-up Structure (Hobs) formation at the pellet rim. At the rim, the burn-up was estimated as 130 MWd/kg UO 2 . After commercial irradiation, PIE was performed at Studsvik. Inner and outer clad oxide thickness measurements were 42 and 5 microns respectively. The measured cold rod diameter varied between 12.20 and 12.25 mm, thus only a small amount of creep-down had occurred from the original diameter of 12.25 mm. Cold gap measurements were taken by diametral compression of the clad onto the fuel. The stiffness changes twice during these measurements, the first (relocated gap) associated with the onset of pellet fragment movement, the second (compressed gap) when the fragments are together and the pellet is compressed. For these rods, the compressed diametral gap was measured as 30 microns. This is in agreement with the pellet and cladding being in contact during the final irradiation cycle, i.e., at ∼12 kW/m. FGR measurements were made after puncturing and values of 2.5%-3.3% were calculated from the extracted gas. The uncertainty is due to different methods of calculation. Ceramography showed a normal crack pattern and no evidence of

  14. Neutronics performances study of silicon carbide as an inert matrix to achieve very high burn-up for light water reactor fuels

    International Nuclear Information System (INIS)

    Chabert, C.; Coulon-Picard, E.; Pelletier, M.

    2007-01-01

    In order to extend the actual limits of light water reactors, the Cea has put emphasis on the exploration of major fuel innovations that would allow us to increase the competitiveness, the safety and flexibility, while keeping the standard PWR environment. Different fuel concepts have been chosen and are actually studied to evaluate their advantages and drawbacks. The objectives of these new fuels are to increase the safety performances and to achieve a very high burn-up. One concept is a CERCER fuel with silicon carbide (SiC) as an inert matrix devoted to reduce the fuel temperature at nominal conditions. Besides the investigation of the neutronic performance, analyses on the thermomechanical performances, the fuel fabrication, the fuel reprocessing and economic aspects have been performed. This paper presents particularly neutronic results obtained for the CERCER fuel. The results show that a very high burn-up, a high safety performance and a better competitiveness cannot be achieved with this fuel concept. (authors)

  15. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

    2008-01-01

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  16. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  17. The effect of dissolved hydrogen on the dissolution of 233U doped UO2(s) high burn-up spent fuel and MOX fuel

    International Nuclear Information System (INIS)

    Carbol, P.; Spahiu, K.

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of 233 U doped UO 2 (s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H 2 pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H 2 pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO 2 , high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10 -6 /yr - 10 -8 /yr with a recommended value of 4x10 -7 /yr for dissolved hydrogen concentrations above 10 -3 M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO 2 and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB

  18. Analysis of UO2 fuel structure for low and high burn-up and its impact on fission gas release

    International Nuclear Information System (INIS)

    Szuta, M.; El-Koliel, M.S.

    1999-01-01

    During irradiation, uranium dioxide (UO 2 ) fuel undergo important restructuring mainly represented by densification and swelling, void migration, equiaxed grain growth, grain subdivision, and the formation of columnar grains. The purpose of this study is to obtain a comprehensive picture of the phenomenon of equiaxed grain growth in UO 2 ceramic material. The change of the grain size in high-density uranium dioxide as a function of temperature, initial grain size, time, and burnup is calculated. Algorithm of fission gas release from UO 2 fuel during high temperature irradiation at high burnup taking into account grain growth effect is presented. Theoretical results are compared with experimental data. (author)

  19. Oxygen stoichiometry shift of irradiated LWR-fuels at high burn-ups: Review of data and alternative interpretation of recently published results

    International Nuclear Information System (INIS)

    Spino, J.; Peerani, P.

    2008-01-01

    The available oxygen potential data of LWR-fuels by the EFM-method have been reviewed and compared with thermodynamic data of equivalent simulated fuels and mixed oxide systems, combined with the analysis of lattice parameter data. Up to burn-ups of 70-80 GWd/tM the comparison confirmed traditional predictions anticipating the fuels to remain quasi stoichiometric along irradiation. However, recent predictions of a fuel with average burn-up around 100 GWd/tM becoming definitely hypostoichiometric were not confirmed. At average burn-ups around 80 GWd/tM and above, it is shown that the fuels tend to acquire progressively slightly hyperstoichiometric O/M ratios. The maximum derived O/M ratio for an average burn-up of 100 GWd/tM lies around 2.001 and 2.002. Though slight, the stoichiometry shift may have a measurable accelerating impact on fission gas diffusion and release

  20. High-Pressure Liquid Chromatography of Irradiated Nuclear Fue - Separation of Neodymium for Burn-up Determination

    DEFF Research Database (Denmark)

    Larsen, N. R.

    1979-01-01

    Neodymium is separated from solutions of spent nuclear fuel by high-pressure liquid chromatography in methanol-nitric acid-water media using an anion-exchange column. Chromatograms obtained by monitoring at 280 nm, illustrate the difficulties especially with the fission product ruthenium in nuclear...

  1. Development of a numerical experimentation method for thermal hydraulics design and evaluation of high burn-up and innovative fuel pins

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Misawa, Takeharu; Baglietto, Emilio; Sorokin, A.P.; Maekawa, Isamu; Ohshima, Hiroyuki; Yamaguchi, Akira

    2003-03-01

    A method of large scale direct numerical simulation of turbulent flows in a high burn-up fuel pin bundle is proposed to evaluate wall shear stress and temperature distributions on the pin surfaces as well as detailed coolant velocity and temperature distributions inside subchannels under various thermal hydraulic conditions. This simulation is aimed at providing a tool to confirm margins to thermal hydraulics design limits of the nuclear fuels and at the same time to be used in design-by-analysis approaches. The method will facilitate thermal hydraulic design of high performance LMFR core fuels characterized by high burn-up, ultra long life, high reliable and safe performances, easiness of operation and maintenance, minimization of radio active wastes, without much relying on such empirical approach as hot spot factor and sub-factors, and above all the high cost mock up experiments. A pseudo direct numerical simulation of turbulence (DNS) code is developed, first on the Cartesian coordinates and then on the curvilinear boundary fit coordinates that enables us to reproduce thermal hydraulics phenomena in such a complicated flow channel as subchannels in a nuclear fuel pin assembly. The coordinate transformation is evaluated and demonstrated to yield correct physical quantities by carrying out computations and comparisons with experimental data with respect to the distributions of various physical quantities and turbulence statistics for fluid flow and heat transfers in various kinds of simple flow channel geometry. Then the boundary fitted pseudo DNS for flows inside an infinite pin array configuration is carried out and compared with available detailed experimental data. In parallel similar calculations are carried out using a commercial code STAR-CD to cross-check the DNS performances. As a results, the pseudo DNS showed reasonable comparisons with experiments as well as the STAR-CD results. Importance of the secondary flow influences is emphasized on the momentum

  2. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  3. Numerical analysis and simulation of behavior of high burn-up PWR fuel pulse-irradiated in reactivity-initiated accident conditions

    International Nuclear Information System (INIS)

    Suzuki, M.; Sugiyama, T.; Udagawa, Y.; Nagase, F.; Fuketa, T.

    2010-01-01

    The four cases of the NSRR experiments, consisting of two room temperature tests and two high temperature tests, using high burn-up PWR fuel rods are analyzed by using the RANNS code to discuss the fuel behavior in hypothetical pulse-irradiation conditions, and the results are compared with metallography observations of ruptured claddings. The cladding rupture occurred by a shear sliding which starts from the tip of incipient crack generated in the hydride dense layer. The analyses reveal that the onset of shear sliding leading to cladding rupture can be closely associated with the stress intensity factor KI at the crack tip and local plastic strain evolution around the tip as well, and that these two factors depend also on the temperature of cladding. Simulation calculations on the basis of experimental conditions reveals that the cladding stress is dependent on the height and half-width of pulse power, and for the same integral enthalpy of pulse a larger half-width mitigates the severity of transient and decreases KI to allow plastic strain by temperature rise, thus failure possibility would be markedly decreased

  4. Long-Term Dry Storage of High Burn-Up Spent Pressurized Water Reactor (PWR) Fuel in TAD (Transportation, Aging, and Disposal) Containers

    International Nuclear Information System (INIS)

    Hwang, Yong Soo

    2008-12-01

    A TAD canister, in conjunction with specially-designed over-packs can accomplish the functions of transportation, aging, and disposal (TAD) in the management of spent nuclear fuel (SNF). Industrial dry cask systems currently available for SNF are licensed for storage-only or for dual-purpose (i.e., storage and transportation). By extending the function to include the indefinite storage and perhaps, eventual geologic disposal, the TAD canister would have to be designed to enhance, among others, corrosion resistance, thermal stability, and criticality-safety control. This investigative paper introduces the use of these advanced iron-based, corrosion-resistant materials for SNF transportation, aging, and disposal.The objective of this investigative project is to explore the interest that KAERI would research and develop its specific SAM coating materials for the TAD canisters to satisfy the requirements of corrosion-resistance, thermal stability, and criticality-controls for long-term dry storage of high burn-up spent PWR fuel

  5. Computational and experimental analysis of causes for local deformation of research reactor U-Mo fuel pin claddings in case of high burn-ups

    International Nuclear Information System (INIS)

    Popov, V.V.; Khmelevsky, M.Ya.; Lukichev, V.A.; Golosov, O.A.

    2005-01-01

    Post-reactor investigations of (U-Mo) fuel pins irradiated in the IVV-2M reactor have allowed to determine: the change in a fuel pin volume; the dimensions and the kind of the local deformation of fuel pin claddings; the amount of gases released under the cladding from the fuel composition, the thickness and appearance of the interaction layer of between the (U-Mo) particles and aluminium as a matrix material. The computational analysis of the stressed-strained state of fuel pins has shown that the major contribution to the increase of the fuel pin volume is made by the fuel swelling caused by the solid products of fission being formed in the process of operation. The emergence of the (U-Mo) fuel-aluminium matrix interaction layers around the (U-Mo) particles results in formation and evolution of lamination cavities inside the fuel composition under the joint action of the pressure of process gases and gaseous fission products. In case of high burn-up a local bulge of a fuel pin cladding is being formed in the fuel lamination area caused by the pressure of gases in the presence of creep in the fuel pin cladding material. The computational results relating to the local strain in a research reactor (U-Mo) fuel pin are in a good accordance with the results of the post-reactor investigations. (author)

  6. Development of methods for burn-up calculations for LWR's

    International Nuclear Information System (INIS)

    Jaschik, W.

    1978-01-01

    This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in which the heterogeneous cell structure and self-shielding effects can explicitly be accounted for. Then the cluster burn-up is simulated by adequate correlation of the burn-up data. The advantage of this method is given by - an exact determination of the real spectrum distribution in the individual fuel element clusters; - an exact determination of the burn-up related spectrum variations for each fuel rod and for each burn-up value obtained; - accounting for heterogeneity of the fuel rod cells and the self-shielding in the fuel; high accuracy of the results of a comparably low effort and - simple handling by largely automating the process of computation. Programed realization was achieved by establishing the RSYST modules ABRAJA, MITHOM, and SIMABB and their implementation within the code system. (orig./HP) [de

  7. A burn-up module coupling to an AMPX system

    International Nuclear Information System (INIS)

    Salvatore Duque, M.; Gomez, S.E.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

    1990-01-01

    The Reactors and Neutrons Division of the Bariloche Atomic Center uses the AMPX system for the study of high conversion reactors (HCR). Such system allows to make neutronic calculations from the nuclear data library (ENDF/B-IV). The Nuclear Engineering career of the Balseiro Institute developed and implemented a burn-up module at a μ-cell level (BUM: Burn-up Module) which agrees with the requirement to be coupled to the AMPX system. (Author) [es

  8. Storage and disposal of high-level radioactive waste from advanced FBR fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2011-01-01

    Waste management of fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P and T) technology was investigated by focusing on thermal constraints due to heat deposition from waste in storage and disposal facilities including economics aspects of those facilities. Partitioning of minor actinides (MAs) and heat-generating fission products in high-level waste can enlarge the containment ratio of waste elements in the glass waste forms and shorten predisposal storage period. Though MAs can be transmuted in FBRs or dedicated transmuters, heat-generating fission products are difficult to be transmuted; they are partitioned and stored for a long time before disposal. The disposal concepts for heat-generating fission products and remainders such as rare-earth elements depend on storage period that ranges from several years to several hundreds of years. Short-term storage results in small size of storage facilities and large size of repositories, and vice versa for long-term storage. This trade-off relation was analyzed by estimating repository size as a function of storage period. The result shows that transmutation of MAs is essentially effective to reduce repository size regardless to storage period, and a combination of P and T can provide a smaller repository than the conventional one by two orders of magnitude. The cost analysis for waste management was also made based on rough assumptions on storage, transportation and repository excluding cost for introducing P and T that are still under evaluation. Cost of waste management for FBR without P and T is 0.25 Yen/kWh that is slightly smaller than that for LWR without P and T, 0.30 Yen/kWh. The introduction of MA transmutation to the FBR results in cost of 0.20 Yen/kWh, and full introduction of P and T provides the smallest cost of 0.08 Yen/kWh. (author)

  9. Inner wall attack and its inhibition method for FBR fuel pin cladding at high burnup

    International Nuclear Information System (INIS)

    Xu Yongli; Long Bin; Li Jingang; Wan Jiaying

    1998-01-01

    The inner wall attack of the modified 316-Ti S.S. cladding tubes manufactured in China used FBR at 10at.% burnup was investigated by means of the out of pile simulation tests. The inner surface morphologies of the cladding tubes attached by fission products Cs, Te, I and Se at 700 deg. C under lower and high oxygen potentials were observed respectively, and the depth of attack was also measured. The burst strength, maximum circum expansion and the appearances of fracture were measured and observed respectively for the cladding tubes attacked by fission products. Based on the mechanism of FBR fuel cladding chemical interaction (FCCI), Cr, Zr and Nb were used as the oxygen absorbers respectively, in order to inhibit the inner wall attack of the cladding tubes. The corrosion morphologies and depth, the penetration depth of the fission products in the inner surface of the cladding tubes were detected. The inhibition effectiveness of the oxygen absorbers for the inner wall attack of the cladding tubes was evaluated. (author)

  10. The CEA-FRAGEMA ramp test programme for the study of the effect of power cycling on PCI at high burn-up

    International Nuclear Information System (INIS)

    Joseph, J.; Atabek, R.; Trotabas, M.

    1983-01-01

    The ramp test programme developed jointly by FRAGEMA and CEA is presented. Today, more than thirty PWR fuel rods have been ramp-tested in experimental reactors SILOE and OSIRIS. Nineteen fuel rods, named 'PRISCA', were base irradiated in BR3 and twelve fuel rods have been refabricated in hot cell by the FABRICE technique. The average fuel burn-up lies between 11 GWd.tU -1 and 46 GWd.tU -1 . In the major cases, the flux profile, during ramp-test, was decentred with respect to the base irradiated flux, and allows to obtain much more information than with a centred flux profile. The failure threshold was established under a set of more than thirty fuel rods of various designs. In particular, the post irradiation examinations allow to locate all rupture locations and thus to define precisely the threshold condition for failure of fuel rods. As an example, the results obtained in the PRISCA 109 experiments are presented. The refabricated fuel rods FABRICE behave identically, with regard to PCI, rather than PRISCA rods. An example of load follow transient in a PWR reactor is presented, and indicates any risk of failure due to PCI. (author)

  11. 3D pin-by-pin power density profiles with high spatial resolution in the vicinity of a BWR control blade tip simulated with coupled neutronics/burn-up calculations

    International Nuclear Information System (INIS)

    Li, J.; Nünighoff, K.; Allelein, H.-J.

    2011-01-01

    Highlights: ► High spatial resolution neutronic and burn-up calculations of quarter BWR fuel element section. ► Coupled MCNP(X)–ORIGEN2.2 simulation using VESTA. ► Control blade history effect was taken into account. ► Determining local power excursion after instantaneous control rod movement. ► Correlation between control blade geometry and occurrence of local power excursions. - Abstract: Pellet cladding interaction (PCI) as well as pellet cladding mechanical interaction (PCMI) are well-known fuel failures in light water reactors, especially in boiling water reactors (BWR). Whereas the thermo-mechanical processes of PCI effects have been intensively investigated in the last decades, only rare information is available on the role of neutron physics. However, each power transient is primary due to neutron physics effects and thus knowledge of the neutron physical background is mandatory to better understand the occurrence of PCI effects in BWRs. This paper will focus on a study of local power excursions in a typical BWR fuel assembly during control rod movements. Burn-up and energy deposition were simulated with high spatial granularity, especially in the vicinity of the control blade tip. It could be shown, that the design of the control blade plays a dominant role for the occurrence of local power peaks while instantaneously moving down the control rod. The main result is, that the largest power peak occurs at the interface between steel handle and absorber rods. A full width half maximum (FWHM) of ±2.5 cm was observed. This means, the local power excursion due to neutron physics phenomena involve approximately five pellets. With the VESTA code coupled MCNP(X)/ORIGEN2.2 calculations were performed with more than 3400 burn-up zones in order to take history effects into account.

  12. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    Matausek, M.

    1984-01-01

    In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

  13. Modelling of thermal mechanical behaviour of high burn-Up VVER fuel at power transients with special emphasis on the impact of fission gas induced swelling of fuel pellets

    International Nuclear Information System (INIS)

    Novikov, V.; Medvedev, A.; Khvostov, G.; Bogatyr, S.; Kuzetsov, V.; Korystin, L.

    2005-01-01

    This paper is devoted to the modelling of unsteady state mechanical and thermo-physical behaviour of high burn-up VVER fuel at a power ramp. The contribution of the processes related to the kinetics of fission gas to the consequences of pellet-clad mechanical interaction is analysed by the example of integral VVER-440 rod 9 from the R7 experimental series, with a pellet burn-up in the active part at around 60 MWd/kgU. This fuel rod incurred ramp testing with a ramp value ΔW 1 ∼ 250 W/cm in the MIR research reactor. The experimentally revealed residual deformation of the clad by 30-40 microns in the 'hottest' portion of the rod, reaching a maximum linear power of up to 430 W/cm, is numerically justified on the basis of accounting for the unsteady state swelling and additional degradation of fuel thermal conductivity due to temperature-induced formation and development of gaseous porosity within the grains and on the grain boundaries. The good prediction capability of the START-3 code, coupled with the advanced model of fission gas related processes, with regard to the important mechanical (residual deformation of clad, pellet-clad gap size, central hole filling), thermal physical (fission gas release) and micro-structural (profiles of intra-granular concentration of the retained fission gas and fuel porosity across a pellet) consequences of the R7 test is shown. (authors)

  14. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Pavelescu, M.; Borza, M.

    1975-01-01

    The authors approach theoretical treatment of isotopic composition changement for nuclear fuel in nuclear reactors. They show the difficulty of exhaustive treatment of burn-up problems and introduce the principal simplifying principles. Due to these principles they write and solve analytically the evolution equations of the concentration for the principal nuclides both in the case of fast and thermal reactors. Finally, they expose and comment the results obtained in the case of a power fast reactor. (author)

  15. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 2: Human reliability analysis and human performance evaluation; Technical issues related to rulemakings; Risk-informed, performance-based initiatives; High burn-up fuel research

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues related to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research

  16. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 2: Human reliability analysis and human performance evaluation; Technical issues related to rulemakings; Risk-informed, performance-based initiatives; High burn-up fuel research

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues related to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  17. Nondestructive, fast methods for burn-up study

    International Nuclear Information System (INIS)

    Schaechter, L.; Hacman, D.; Mot, O.

    1977-01-01

    Nondestructive methods, based on high resolution-spectrometry successfully applied at Institute for Atomic Physics are presented. These methods are preferred to destructive chemical methods; the latter being costly and lengthy and not suitable for statistical prediction of nuclear fuel behaviour. The following methods are developed: methods for determining the burn up of fuel elements and fuel assemblies; a method for determining the U 235 and Pu 239 contributions to the burn up and a code written in FORTRAN IV for numerical calculation of Pu 239 fission vs. burn up; a high precision method for burnup determination by adding burnable poison; a method for prediction of specific power distribution in the fuel elements of a research or power reactors; a method for determining the power output of the fuel element in an operating power reactor; a method for determining the content of Pu 239 of the fuel element irradiated in a reactor. The results which were obtained by these methods improved the fuel management at the VVR-S reactor at Institute for Atomic Physics, Bucharest and may be applied to other reactor types [fr

  18. Determination of burn-up of irradiated nuclear fuels using mass spectrometry

    International Nuclear Information System (INIS)

    Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

    2017-01-01

    Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

  19. Determination of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Kristak, J.; Vobecky, M.

    1973-01-01

    Samples containing a known content of 235 U were irradiated with several different neutron doses and activities were determined of radionuclides including 125 Sb, 144 Ce, 134 Cs, 154 Eu, 103 Ru, 95 Zr. The values thus obtained were divided by the 137 Cs activity value. The resulting neutron dose-dependent value is plotted into a calibration graph. The degree of nuclear fuel burn-up is obtained from the graph using an experimentally determined ratio of the activities of the above radionuclides. (B.S.)

  20. Recent observations at the post-irradiation examination of low-enriched U-Mo miniplates irradiated to high burn-up

    International Nuclear Information System (INIS)

    Hofman, G.L.; Kim, Y.S.; Finlay, M.R.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.; Clark, C.R.

    2003-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL Alpha-Gamma Hot Cell Facility. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 6.5 wt% to 10 wt% molybdenum. In addition, two miniplates contained solid U-10wt%Mo foils. Recent results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from a companion test performed to 50% burnup, RERTR-5, are presented. (author)

  1. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  2. Establishing a PWR burn-up library

    International Nuclear Information System (INIS)

    Lutz, D.C.

    1981-01-01

    Starting out from data file ENDF/B IV /1/, a cross-section library has been established for the calculation of operating conditions in pressurized water reactors of the type used in BIBLIS B. The library includes macroscopic, homogenized 2-group cross-sections for all types of fuel elements used in this reactor, including those equipped with boron glass rods. For their calculation the previous irradiation of the fuel has been taken into consideration by approximation. Information on fuel consumption from cell burn-up calculations has been stored in a separate data file. It was designed as a base for the determination of cross sections to be used in the calculation of the incident ''main-steam pipe fracture''. For this library the description of cross sections as a function of the moderator status chose the water densities at 300 0 C/155 bar, 190 0 C/140 bar and 100 0 C/100 bar as fixed values. The burn-up library has been tested by a three-dimensional calculation for the 1sup(st) cycle of the BIBLIS B-reactor using program QUABOX /2/. This showed variances with the anticipated course concerning critically, which can be explained almost quantitatively by known deficiencies of the ENDF/b-IV library. (orig.) [de

  3. Criterion for burn-up conditions in gas-cooled cryogenic current leads

    International Nuclear Information System (INIS)

    Bejan, A.; Cluss, E.M. Jr.

    1976-01-01

    Superconducting magnets are energized through helium vapour-cooled cryogenic current leads operating at high ratios of current to mass flow. The high current operation where lead temperature, runaway, and eventual burn-up are likely to occur is investigated. A simple criterion for estimating the burn-up operation conditions (current, mass flow) for a given lead geometry (cross-sectional area, length, heat exchanger area) is presented. This article stresses the role played by the available heat exchanger area in avoiding burn-up at high ratios of current to mass flow. (author)

  4. Development of continuous energy Monte Carlo burn-up calculation code MVP-BURN

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Nakagawa, Masayuki; Sasaki, Makoto

    2001-01-01

    Burn-up calculations based on the continuous energy Monte Carlo method became possible by development of MVP-BURN. To confirm the reliably of MVP-BURN, it was applied to the two numerical benchmark problems; cell burn-up calculations for High Conversion LWR lattice and BWR lattice with burnable poison rods. Major burn-up parameters have shown good agreements with the results obtained by a deterministic code (SRAC95). Furthermore, spent fuel composition calculated by MVP-BURN was compared with measured one. Atomic number densities of major actinides at 34 GWd/t could be predicted within 10% accuracy. (author)

  5. Core burn-up calculation method of JRR-3

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Yamashita, Kiyonobu

    2007-01-01

    SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)

  6. Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

    2003-04-01

    A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

  7. Observations on the CANDLE burn-up in various geometries

    International Nuclear Information System (INIS)

    Seifritz, W.

    2007-01-01

    We have looked at all geometrical conditions under which an auto catalytically propagating burnup wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning 'spaghetti' reactor and the azimuthally burning 'pancake' reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometry, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed. A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology

  8. Numerical solution of matrix exponential in burn-up equation using mini-max polynomial approximation

    International Nuclear Information System (INIS)

    Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

    2015-01-01

    Highlights: • We propose a new numerical solution of matrix exponential in burn-up depletion calculations. • The depletion calculation with extremely short half-lived nuclides can be done numerically stable with this method. • The computational time is shorter than the other conventional methods. - Abstract: Nuclear fuel burn-up depletion calculations are essential to compute the nuclear fuel composition transition. In the burn-up calculations, the matrix exponential method has been widely used. In the present paper, we propose a new numerical solution of the matrix exponential, a Mini-Max Polynomial Approximation (MMPA) method. This method is numerically stable for burn-up matrices with extremely short half-lived nuclides as the Chebyshev Rational Approximation Method (CRAM), and it has several advantages over CRAM. We also propose a multi-step calculation, a computational time reduction scheme of the MMPA method, which can perform simultaneously burn-up calculations with several time periods. The applicability of these methods has been theoretically and numerically proved for general burn-up matrices. The numerical verification has been performed, and it has been shown that these methods have high precision equivalent to CRAM

  9. Oxide fuel fabrication technology development of the FaCT project (5). Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

    International Nuclear Information System (INIS)

    Ohtsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

    2011-01-01

    This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production, both of which are key points for the commercialized use of ODS steels as long-life fuel cladding tubes. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed, where the handling of elemental powder is prohibited in the process. (author)

  10. Burn-up measurements coupling gamma spectrometry and neutron measurement

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H.; Pin, P. [AREVA/CANBERRA, 1 rue des Herons, 78182 St Quentin-en-Yvelines Cedex (France); Lebrun, A. [IAEA, Wagramer Strasse 5, PO Box 100, Vienna (Austria); Oriol, L.; Saurel, N. [CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Gain, T. [AREVA/COGEMA Reprocessing Business Unit, La Hague, 50444 Beaumont Hague Cedex (France)

    2006-07-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  11. Burn-up measurements coupling gamma spectrometry and neutron measurement

    International Nuclear Information System (INIS)

    Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.

    2006-01-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  12. Approach to lithium burn-up effect in lithium ceramics

    International Nuclear Information System (INIS)

    Rasneur, B.

    1994-01-01

    The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

  13. Deuterides of light elements: low-temperature thermonuclear burn-up and applications to thermonuclear fusion problems

    International Nuclear Information System (INIS)

    Frolov, A.M.; Smith, V.H.; Smith, G.T.

    2002-01-01

    Thermonuclear burn-up and thermonuclear applications are discussed for a number of deuterides and DT hydrides of light elements. These deuterides and corresponding DT hydrides are often used as thermonuclear fuels or components of such fuels. In fact, only for these substances thermonuclear energy gain exceeds (at some densities and temperatures) the bremsstrahlung loss and other high-temperature losses, i.e., thermonuclear burn-up is possible. Herein, thermonuclear burn-up in these deuterides and DT hydrides is considered in detail. In particular, a simple method is proposed to determine the critical values of the burn-up parameter x c for these substances and their mixtures at different temperatures and densities. The results for equimolar DT mixtures coincide quite well with the results of previous calculations. Also, the natural or Z limit is determined for low-temperature thermonuclear burn-up in the deuterides of light elements. (author)

  14. Two dimensional burn-up calculation of TRIGA core

    International Nuclear Information System (INIS)

    Persic, A.; Ravnik, M.; Slavic, S.

    1996-01-01

    TRIGLAV is a new computer program for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport program WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. In this paper the calculation model of diffusion constants and burn-up calculation are described and some results of calculations for TRIGA MARK II reactor are presented. (author)

  15. Estimation of the impact of manufacturing tolerances on burn-up calculations using Monte Carlo techniques

    Energy Technology Data Exchange (ETDEWEB)

    Bock, M.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, Garching (Germany). Forschungszentrum

    2012-11-01

    In recent years, the availability of computing resources has increased enormously. There are two ways to take advantage of this increase in analyses in the field of the nuclear fuel cycle, such as burn-up calculations or criticality safety calculations. The first possible way is to improve the accuracy of the models that are analyzed. For burn-up calculations this means, that the goal to model and to calculate the burn-up of a full reactor core is getting more and more into reach. The second way to utilize the resources is to run state-of-the-art programs with simplified models several times, but with varied input parameters. This second way opens the applicability of the assessment of uncertainties and sensitivities based on the Monte Carlo method for fields of research that rely heavily on either high CPU usage or high memory consumption. In the context of the nuclear fuel cycle, applications that belong to these types of demanding analyses are again burn-up and criticality safety calculations. The assessment of uncertainties in burn-up analyses can complement traditional analysis techniques such as best estimate or bounding case analyses and can support the safety analysis in future design decisions, e.g. by analyzing the uncertainty of the decay heat power of the nuclear inventory stored in the spent fuel pool of a nuclear power plant. This contribution concentrates on the uncertainty analysis in burn-up calculations of PWR fuel assemblies. The uncertainties in the results arise from the variation of the input parameters. In this case, the focus is on the one hand on the variation of manufacturing tolerances that are present in the different production stages of the fuel assemblies. On the other hand, uncertainties that describe the conditions during the reactor operation are taken into account. They also affect the results of burn-up calculations. In order to perform uncertainty analyses in burn-up calculations, GRS has improved the capabilities of its general

  16. Burn-up measurement in the HTR-module-reactor

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

    The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

  17. TRIGA criticality experiment for testing burn-up calculations

    International Nuclear Information System (INIS)

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    1999-01-01

    A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)

  18. Scenario study on the FBR deployment

    International Nuclear Information System (INIS)

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

    2000-12-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  20. Increased fuel burn-up and fuel cycle equilibrium

    International Nuclear Information System (INIS)

    Debes, M.

    2001-01-01

    Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

  1. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

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

  2. Nuclear fuel burn-up economy; Ekonomija izgaranja nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-07-01

    In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

  3. Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

    International Nuclear Information System (INIS)

    Garcia-Herranz, Nuria; Cabellos, Oscar; Sanz, Javier; Juan, Jesus; Kuijper, Jim C.

    2008-01-01

    Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files

  4. Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

    2008-04-15

    Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

  5. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  6. High temperature structural design and R and Ds for heat transport system components of FBR 'Monju'

    International Nuclear Information System (INIS)

    Sumikawa, Masaharu; Nakagawa, Yukio; Fukuda, Yoshio; Sukegawa, Masayuki; Ishizaki, Tairo.

    1980-01-01

    The machines and equipments of cooling system for the fast breeder prototype reactor ''Monju'' are operated in creep temperature region, and the upper limit temperature to apply the domestic structural design standard for nuclear machines and equipment is exceeded, therefore the guideline for high temperature structural design is being drawn up, reflecting the results of recent research and development, by the Power Reactor and Nuclear Fuel Development Corp. and others. In order to obtain the basic data for the purpose, the tests on the high temperature characteristics of main structural members and structural elements were carried out, and eight kinds of the inelastic structural analysis program ''HI-EPIC'' series were developed, thus the fundamental technologies of structural desigh in non-linear region were established. Also in the non-linear region, enormous physical quantities must be evaluated, and in the design method based on real elastic analysis, many design diagrams must be employed, therefore for the purpose of improving the reliability of evaluation, the automatic evaluation program ''HI-TEP'' was developed, and preparation has been made for the design of actual machines. The high temperature structural design in ''Monju'', the development of inelastic structural analysis program and high temperature structural analysis evaluation program, and the development of high temperature structures and materials are described. (Kako, I.)

  7. Universal high-temperature heat treatment furnace for FBR mixed uranium and plutonium carbide fuel

    International Nuclear Information System (INIS)

    Handa, Muneo; Takahashi, Ichiro; Watanabe, Hitoshi

    1978-10-01

    A universal high-temperature heat treatment furnace for LMFBR advanced fuels was installed in Plutonium Fuel Laboratory, Oarai Research Establishment. Design, construction and performance of the apparatus are described. With the apparatus, heat treatment of the fuel under a controlled gas atmosphere and quenching of the fuel with blowing helium gas are possible. Equipment to measure impurity gas release of the fuel is also provided. Various plutonium enclosure techniques, e.g., a gas line filter with new exchange mechanics, have been developed. In performance test, results of the enclosure techniques are described. (author)

  8. Enlarged Halden programme group meeting on high burn-up fuel performance, safety and reliability and degradation of in-core materials and water chemistry effects and man-machine systems research. Volume II

    International Nuclear Information System (INIS)

    1999-01-01

    Academy of Sciences, KFKI Atomic Energy Research Institute, the N.V. KEMA, the Netherlands, the Russian Research Centre 'Kurchatov Institute', the Slovakian VUJE - Nuclear Power Plant Research Institute, and from USA: the ABB Combustion Engineering Inc., the Electric Power Research Institute (EPRI), and the General Electric Co. The right to utilise information originating from the research work of the Halden Project is limited to persons and undertakings specifically given this right by one of these Project member organisations. The activities in the area of fuel and materials performance are based on extensive in-reactor measurements. The programmes are expanding in the areas of fuel performance at extended burn-ups, waterside corrosion and material testing in general. Development of in-core instruments is an important activity in support of the experimental programmes. The research programme at the Halden Project addresses the research needs of the nuclear industry in connection with introduction of digital I and C systems in NPPs. The programme provides information supporting design and licensing of upgraded, computer-based control room systems, and demonstrates the benefits of such systems through validation experiments in Halden's experimental research facility, HAMMLAB and pilot installations in NPPs. The Enlarged Halden Programme Group Meeting at Loen, Norway, was arranged to provide an opportunity to present results of work carried out at Halden and within participating organisations, and to encourage comments and impulses related to future Halden Project work. This HPR-351 relates to the fuel and materials part of the meeting and is divided in two volumes, HPR-351 Volume I and HPR-351 Volume II. The corresponding collection of papers in the man-machine area are given in one volume, HPR-352 Volume I. The overall programme of the Loen Enlarged Meeting covering the Fuel and Materials Research is given in the following pages. The papers with denomination HWR have

  9. Creep-fatigue life property of FBR high-temperature structural materials under tension-torsion loading and life evaluation method

    International Nuclear Information System (INIS)

    Ogata, Takashi; Nitta, Akito

    1994-01-01

    Creep-fatigue damage in high temperature structural components in a FBR progress under multiaxial stress condition depending on their operating conditions and configuration. Therefore, multiaxial stress effects on creep-fatigue damage evolution must be clarified to make precise creep-fatigue damage evaluation of these components. In this study, creep-fatigue tests in FBR high temperature materials such as SUS304, 316FR stainless steels and a modified 9Cr steel were conducted under biaxial stress subjecting tension-compression and torsion loading, in order to examine biaxial stress effects on failure mechanism and life property, and to discuss creep-fatigue life evaluation methods under biaxial stress. Main results obtained in this study are summarized as follows: 1. The main cracks under cyclic torsion loading propagated by shear mode in three materials. But intergranular failure was occurred in SUS304 and 316FR, and transgranular failure was observed in Mod.9Cr steel. 2. Nonlinear damage accumulation model proposed based on uniaxial creep-fatigue test results was extended to apply for creep-fatigue damage evaluation under biaxial stress state by considering the biaxial stress effects on fatigue and creep damage evolution. 3. It was confirmed that creep-fatigue life under biaxial stress could be predicted by the extended evaluation method with higher accuracy than existing methods. (author)

  10. CRISTAL V1: Criticality package for burn up credit calculations

    International Nuclear Information System (INIS)

    Gomit, Jean-Michel; Cousinou, Patrick; Gantenbein, Francoise; Diop, Cheikh; Fernandez de Grado, Guy; Mijuin, Dominique; Grouiller, Jean-Paul; Marc, Andre; Toubon, Herve

    2003-01-01

    The first version of the CRISTAL package, created and validated as part of a joint project between IRSN, COGEMA and CEA, was delivered to users in November 1999. This fruitful cooperation between IRSN, COGEMA and CEA has been pursued until 2003 with the development and the validation of the package CRISTAL V1, whose main objectives are to improve the criticality safety studies including the Burn up Credit effect. (author)

  11. Calculation of triton confinement and burn-up in tokamaks

    International Nuclear Information System (INIS)

    Anderson, D.; Battistoni, P.

    1987-01-01

    An analytical investigation is made of the confinement and subsequent burn-up of fusion produced tritons in a deuterium Tokamak plasma. Explicit approximations are obtained for the triton confinement factor, clearly displaying the scaling with physical parameters. The importance of pitch angle scattering losses during the triton slowing down is also estimated. A comparison with experiments and numerical calculations on the FT Tokamak slows good qualitative agreement. (authors)

  12. A Study for Burn-up Calculation applied on 400MWth PBMR Core

    International Nuclear Information System (INIS)

    Luu, Nam Hai; Kim, Hong Chul; Kim, Soon Young; Kim, Jong Kyung; Noh, Jae Man

    2007-01-01

    The 400MWth Pebble-bed Modular Reactor (PBMR) is an advanced high temperature gas cooled-reactor (HTGR). It possesses a very high efficiency and attractive economics without compromising the high levels of passive safety expected of advanced nuclear designs. With this reason, PBMR is a target which researchers especially in nuclear engineering field study carefully and therefore it is regarded as the leader in the power generation field. There are many research results about benchmark problems but results of the burn-up process are still poor. Hence, in this study a burn-up calculation was performed with PBMR using MONTEBURNS code in which MCNP modeling linked a depletion systems is used

  13. Three dimensional Burn-up program parallelization using socket programming

    International Nuclear Information System (INIS)

    Haliyati R, Evi; Su'ud, Zaki

    2002-01-01

    A computer parallelization process was built with a purpose to decrease execution time of a physics program. In this case, a multi computer system was built to be used to analyze burn-up process of a nuclear reactor. This multi computer system was design need using a protocol communication among sockets, i.e. TCP/IP. This system consists of computer as a server and the rest as clients. The server has a main control to all its clients. The server also divides the reactor core geometrically to in parts in accordance with the number of clients, each computer including the server has a task to conduct burn-up analysis of 1/n part of the total reactor core measure. This burn-up analysis was conducted simultaneously and in a parallel way by all computers, so a faster program execution time was achieved close to 1/n times that of one computer. Then an analysis was carried out and states that in order to calculate the density of atoms in a reactor of 91 cm x 91 cm x 116 cm, the usage of a parallel system of 2 computers has the highest efficiency

  14. Development of hydrogen production technology using FBR

    International Nuclear Information System (INIS)

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

    2004-06-01

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

  15. Burn-up TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

    Persic, A.; Ravnik, M.; Zagar, T.

    1998-01-01

    Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

  16. Establishing the fuel burn-up measuring system for 106 irradiated assemblies of Dalat reactor by using gamma spectrometer method

    International Nuclear Information System (INIS)

    Nguyen Minh Tuan; Pham Quang Huy; Tran Tri Vien; Trang Cao Su; Tran Quoc Duong; Dang Tran Thai Nguyen

    2013-01-01

    The fuel burn-up is an important parameter needed to be monitored and determined during a reactor operation and fuel management. The fuel burn-up can be calculated using computer codes and experimentally measured. This work presents the theory and experimental method applied to determine the burn-up of the irradiated and 36% enriched VVR-M2 fuel type assemblies of Dalat reactor. The method is based on measurement of Cs-137 absolute specific activity using gamma spectrometer. Designed measuring system consists of a collimator tube, high purity Germanium detector (HPGe) and associated electronics modules and online computer data acquisition system. The obtained results of measurement are comparable with theoretically calculated results. (author)

  17. Simulation of triton burn-up in JET plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, M J; Balet, B; Jarvis, O N; Stubberfield, P M [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

  18. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems

    Directory of Open Access Journals (Sweden)

    Thomas Frosio

    2017-01-01

    Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.

  19. Technical description of the burn-up software system MOP

    International Nuclear Information System (INIS)

    Schutte, C.K.

    1991-05-01

    The burn-up software system MOP is a research tool primary intended to study the behaviour of fission products in any reactor composition. Input data are multi-group cross-sections and data concerning the nuclide chains. An option is available to calculate a fundamental mode neutron spectrum for the specified reactor composition. A separate program can test the consistency of the specified nuclide chains. Options are available to calculate time-dependent cross-sections of lumped fission products and to take account of the leakage of gaseous fission products from the reactor core. The system is written in FORTRAN77 for a CYBER computer, using the operating system NOS/BE. The report gives a detailed technical description of the applied algorithms and the flow and storage of data. Information is provided for adapting the system to other computer configurations. (author). 5 refs.; 11 figs

  20. Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

    International Nuclear Information System (INIS)

    Tohjoh, Masayuki; Endo, Tomohiro; Watanabe, Masato; Yamamoto, Akio

    2006-01-01

    As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore, if statistical error alone is evaluated, the errors in Monte Carlo burn-up calculations may be underestimated. To make clear this effect of error propagation on Monte Carlo burn-up calculations, we here proposed an equation that can predict the variance of nuclide number densities after burn-up calculations, and we verified this equation using enormous numbers of the Monte Carlo burn-up calculations by changing only the initial random numbers. We also verified the effect of the number of burn-up calculation points on Monte Carlo burn-up calculations. From these verifications, we estimated the errors in Monte Carlo burn-up calculations including both statistical and propagated errors. Finally, we made clear the effects of error propagation on Monte Carlo burn-up calculations by comparing statistical errors alone versus both statistical and propagated errors. The results revealed that the effects of error propagation on the Monte Carlo burn-up calculations of 8 x 8 BWR fuel assembly are low up to 60 GWd/t

  1. FBR/VHTR deployment scenarios in Japan

    International Nuclear Information System (INIS)

    Richards, Matt; Kunitomi, Kazuhiko

    2008-01-01

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

  2. Effect of burn-up on the thermal conductivity of uranium dioxide up to 100.000 MWd t-1

    International Nuclear Information System (INIS)

    Ronchi, C.; Sheindlin, M.; Staicu, D.; Kinoshita, M.

    2004-01-01

    The thermal diffusivity and specific heat of reactor-irradiated UO 2 fuel have been measured. Starting from end-of-life conditions at various burn-ups, measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage were determined. In this context, particular emphasis was given to the behaviour of samples displaying the high burn-up rim structure. Recovery stages could be thoroughly investigated in samples that were irradiated at low burn-ups and/or at high irradiation temperatures. Other samples, in particular those exhibiting the characteristic rim structure, disintegrated at temperatures slightly higher than the irradiation temperature. Finally, from a database of several thousand measurements, an accurate formula for the in-pile thermal conductivity of UO 2 up to 100 GWd t -1 was developed, taking into account all the relevant effects and structural changes induced by reactor burn-up

  3. Calculation of isotope burn-up and change in efficiency of absorbing elements of WWER-1000 control and protection system during burn-up

    International Nuclear Information System (INIS)

    Timofeeva, O.A.; Kurakin, K.U.

    2006-01-01

    The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)

  4. Study on the sensitivity of Self-Powered Neutron Detectors (SPND) and its change due to burn-up

    International Nuclear Information System (INIS)

    Cho, Gyuseong; Lee, Wanno; Yoon, Jeong-Hyoun.

    1996-01-01

    Self-Powered Neutron Detectors (SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with its usage, it has some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation was accomplished to calculate the escape probability as a function of the birth position for the typical geometry of rhodium-based SPNDs. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity is calculated as a function of burn-up time in the reactor. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long-term usage. (author)

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

    International Nuclear Information System (INIS)

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

    2004-03-01

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

  6. Fundamental burn-up mode in a pebble-bed type reactor

    International Nuclear Information System (INIS)

    Chen, Xue-Nong; Kiefhaber, Edgar; Maschek, Werner

    2008-01-01

    This paper deals with a pebble-bed type reactor, in which the fuel is loaded from one side (top) and discharged from the other side (bottom). A boundary value problem of a single group diffusion equation coupled with simplified burn-up equations is studied, where the natural radioactive decay processes are neglected in the burn-up modelling. An asymptotic burning wave solution is found analytically in the one-dimensional case, which is called as fundamental burn-up mode. Among this solution family there are two particular cases, namely, a classic fundamental solution with a zero burn-up and a partial solitary burn-up wave solution with a highest burn-up. An example of Th-U conversion is considered and the solutions are presented in order to show the mechanism of the burning wave. (author)

  7. Comparison of measured and calculated burn-up of AVR-Fuel-Elements

    Energy Technology Data Exchange (ETDEWEB)

    Wagemann, R.

    1974-03-15

    Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

  8. Effect of local burn-up variation on computed mean nuclide concentrations

    International Nuclear Information System (INIS)

    Moeller, W.

    1982-01-01

    Mean concentrations of U-235, U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 in some volume areas of WWER-440 fuel assemblies have been calculated from corresponding burn-up microdistribution data and compared with those calculated from burn-up mean values. Differences occurring were below 3% for the uranium nuclides but, at low burn-ups, considerable for Pu-241 and Pu-242. (author)

  9. Reactor structure for FBR

    International Nuclear Information System (INIS)

    Yamazaki, Tsukasa.

    1986-01-01

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

  10. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  11. Structural dynamics in FBR

    International Nuclear Information System (INIS)

    Bhoje, S.B.

    2003-01-01

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

  12. FBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Tsugio.

    1986-01-01

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

  13. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

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

  14. Fuel burn-up distribution and transuranic nuclide contents produced at the first cycle operation of AP1000

    International Nuclear Information System (INIS)

    Jati Susilo; Jupiter Sitorus Pane

    2016-01-01

    AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO 2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB 2 , Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and produce energy, fission products and new neutron. Because of the U-238 neutron absorption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic cross section, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU. Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. (author)

  15. Studies on the dissolution of mixed oxide spent fuel from FBR

    International Nuclear Information System (INIS)

    Nemoto, Shin-ichi; Shibata, Atsuhiro; Shioura, Takao; Okamoto, Fumitoshi; Tanaka, Yasumasa

    1995-01-01

    At the Chemical Processing Facility(CPF) in the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation(PNC), since 1982 Laboratory scale hot experiments have been carried out on the development of reprocessing technology for FBR mixed oxide fuel. The spent fuel pins which have been used in out experiments were irradiated in Experimental Fast Reactor 'Joyo' Phenix (France) and DFR(UK). Burn-up of the fuel pins were 4,400-100,000 MWd/t. This paper Summarizes a dissolution study that have been performed to define the Key parameters affecting dissolution rate such as concentration of nitric acid, burn-up, and temperature. And this paper also discusses about the character of releasing 85 Kr in chopping and dissolution process, and about the amount of insoluble residue. (author)

  16. Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.

    1996-01-01

    The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)

  17. Analysis on burn-up behaviors for accelerator-driven sub-critical facility

    International Nuclear Information System (INIS)

    Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qinbiao; Ding Dazhao

    2000-01-01

    An analysis is performed on burn-up behaviors for accelerator-driven sub-critical reactor by means of the code PASC-1 for neutronics calculation, the code CBURN for burn-up calculation and 44 group constants is processed by CENDL-2 and ENDF/B-6 using NJOY-91.91

  18. Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Rodriguez Rivada, A.

    2014-07-01

    Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)

  19. Determination of nuclear fuel burn-up using mass spectrometric techniques

    International Nuclear Information System (INIS)

    Saha, B.; Bagyalakshmi, R.; Periaswami, G.; Kavimandan, V.D.; Chitambar, S.A.; Jain, H.C.; Mathews, C.K.

    1977-01-01

    Determination of burn-up using a stable fission product monitor such as 148 Nd and heavy elements, determined by isotope dilution mass spectrometry gives the most accurate data. This report describes the work carried out to standardise the conditions for burn-up determination. Some typical results are given. (author)

  20. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

  1. Evaluation of the commercial FBR introduction date

    International Nuclear Information System (INIS)

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

    1981-09-01

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

  2. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  3. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    International Nuclear Information System (INIS)

    Barkauskas, V.; Plukiene, R.; Plukis, A.

    2016-01-01

    Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

  4. Research and development of FBR fuel reprocessing in PNC

    International Nuclear Information System (INIS)

    Hoshino, T.

    1976-05-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  6. Application of seismic isolation technology to demonstration FBR

    International Nuclear Information System (INIS)

    Kato, Muneaki

    1994-01-01

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

  7. Study on coated layer material performance of coated particle fuel FBR (2). High temperature property and capability of coating to thick layer of TiN

    International Nuclear Information System (INIS)

    Naganuma, Masayuki; Mizuno, Tomoyasu

    2002-08-01

    'Helium Gas Cooled Coated Particle Fuel FBR' is one of attractive core concepts in the Feasibility Study on Commercialized Fast Reactor Cycle System in Japan, and the design study is presently proceeded. As one of key technologies of this concept, the coated layer material is important, and ceramics is considered to be a candidate material because of the superior refractory. Based on existing knowledge, TiN is regarded to be a possible candidate material, to which some property tests and evaluations have been conducted. In this study, preliminary tests about the high temperature property and the capability of thick layer coating of TiN have been conducted. Results of these tests come to the following conclusions. Heating tests of two kinds of TiN layer specimens coated by PVD (Physical Vapor Deposition) and CVD (Chemical Vapor Deposition) were conducted. As a result, as for CVD coating specimens, remarkable charge was not observed on the layer up to 2,000degC, therefore we concluded that the layer by CVD had applicability up to high temperature of actual operation level. On the other hand, as for PVD coating specimens, an unstable behavior that the layer changed to a mesh like texture was observed on a 2,000degC heated specimen, therefore the applied PVD method is not considered to be promising as the coating technique. The surface conditions of some parts inside CVD device were investigated in order to evaluate possibility of TiN thick coating (∼100 μm). As a result, around 500 μm of TiN coating layer was observed on the condition of multilayer. Therefore, we conclude that CVD has capability of coating up to thick layer in actual coated particle fuel fabrication. (author)

  8. Fast-ion diffusion measurements from radial triton burn up studies

    International Nuclear Information System (INIS)

    McCauley, J.S.; Budny, R.; McCune, D.; Strachan, J.D.

    1993-08-01

    A fast-ion diffusion coefficient of 0.1 ± 0.1 m 2 s -1 has been deduced from the triton burnup neutron emission profile measured by a collimated array of helium-4 spectrometers. The experiment was performed with high-power deuterium discharges produced by Princeton University's Tokamak Fusion Test Reactor (TFTR). The fast ions monitored were the 1.0 MeV tritons produced from the d(d,t)p. These tritons ''burn up'' with deuterons and emit a 14 MeV neutron by the d(t,α)n reaction. The ratio of the measured to calculated DT yield is typically 70%. The measured DT profile width is comparable to that predicted by the TRANSP transport code during neutral beam heating and narrower after the beam heating ended

  9. Application of reactivity method to MTR fuel burn-up measurement

    International Nuclear Information System (INIS)

    Zuniga, A.; Ravnik, M.; Cuya, R.

    2001-01-01

    Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

  10. Full Core Burn-up Calculation at JRR-3 with MVP-BURN

    International Nuclear Information System (INIS)

    Komeda, Masao; Yamamoto, Kazuyoshi; Kusunoki, Tsuyoshi

    2008-01-01

    Research reactors use a burnable poison to suppress an excess reactivity in the beginning of reactor lifetime. The JRR-3 (Japan Research Reactor No.3) has used cadmium wires of radius 0.02 cm as a burnable poison. This report describes burn-up calculations of plate fuel models and full core models with MVP-BURN, which is a burn-up calculation code using Monte Carlo method and has been developed in JAEA (Japan Atomic Energy Agency). As the results of calculations of plate models, between a model composed of one burn-up region along the radius direction and a model composed of a few burn-up regions along the radius direction, the effective absorption cross section of 113 Cd has had different tendency on reaching approximate 40. day (10000 MWd/t). And as results of calculations of full core model, it has been indicated that k eff is almost same till approximate 80. day (22000 MWd/t) between a model composed of one burn-up region along the vertical direction and a model composed of a few burn-up regions along the vertical direction. However difference of 113 Cd burn-up becomes pronounced and each k eff makes a difference after 80. day. (authors)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  12. FBR and RBR particle bed space reactors

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. Effect of burn-up on the radioactivation behavior of cladding hull materials studied using the ORIGEN-S code

    International Nuclear Information System (INIS)

    Min Ku Jeon; Chang Hwa Lee; Jung Hoon Choi; In Hak Cho; Kweon Ho Kang; Hwan-Seo Park; Geun Il Park; Chang Je Park

    2013-01-01

    The effect of fuel burn-up on the radioactivation behavior of cladding hull materials was investigated using the ORIGEN-S code for various materials of Zircaloy-4, Zirlo, HANA-4, and HANA-6 and for various fuel burn-ups of 30, 45, 60, and 75 GWD/MTU. The Zircaloy-4 material is the only one that does not contain Nb as an alloy constituent, and it was revealed that 125 Sb, 125m Te, and 55 Fe are the major sources of radioactivity. On the other hand, 93m Nb was identified as the most radioactive nuclide for the other materials although minor radioactive nuclides varied owing to their different initial constituents. The radioactivity of 94 Nb was of particular focus owing to its acceptance limit against a Korean intermediate-/low-level waste repository. The radioactivation calculation results revealed that only Zircaloy-4 is acceptable for the Korean repository, while the other materials required at least 4,900 of Nb decontamination factor owing to the high radioactivity of 94 Nb regardless of the fuel burn-up. A discussion was also made on the feasibility of Zr recovery methods (chlorination and electrorefining) for selective recovery of Zr so that it can be disposed of in the Korean repository. (author)

  14. Technical development on burn-up credit for spent LWR fuels

    International Nuclear Information System (INIS)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  15. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  16. Calculation of heat rating and burn-up for test fuel pins irradiated in DR 3

    International Nuclear Information System (INIS)

    Bagger, C.; Carlsen, H.; Hansen, K.

    1980-01-01

    A summary of the DR 3 reactor and HP1 rig design is given followed by a detailed description of the calculation procedure for obtaining linear heat rating and burn-up values of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially regarding features like end pellet contribution to power as a function of burn-up, gamma heat contributions, and evaluation of local values of heat rating and burn-up. Included in the report is also a description of the fast flux- and cladding temperature calculation techniques currently used. A good agreement between measured and calculated local burn-up values is found. This gives confidence to the detailed treatment of the data. (author)

  17. Experimental studies of spent fuel burn-up in WWR-SM reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alikulov, Sh. A.; Baytelesov, S.A.; Boltaboev, A.F.; Kungurov, F.R. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan); Menlove, H.O.; O’Connor, W. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Osmanov, B.S., E-mail: bari_osmanov@yahoo.com [Research Institute of Applied Physics, Vuzgorodok, 100174 Tashkent (Uzbekistan); Salikhbaev, U.S. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan)

    2014-10-01

    Highlights: • Uranium burn-up measurement from {sup 137}Cs activity in spent reactor fuel. • Comparison to reference sample with known burn-up value (ratio method). • Cross-check of the approach with neutron-based measurement technique. - Abstract: The article reports the results of {sup 235}U burn-up measurements using {sup 137}Cs activity technique for 12 nuclear fuel assemblies of WWR-SM research reactor after 3-year cooling time. The discrepancy between the measured and the calculated burn-up values was about 3%. To increase the reliability of the data and for cross-check purposes, neutron measurement approach was also used. Average discrepancy between two methods was around 12%.

  18. The build-up and characterization of nuclear burn-up wave in a fast ...

    Indian Academy of Sciences (India)

    K V Anoop

    2018-02-07

    Feb 7, 2018 ... evaluating the quality of the wave by the researchers working in the field of nuclear burn-up wave build-up and propagation. Keywords. ... However, there are concerns relating to the nuclear safety, ... Simulation studies have.

  19. The cluster burn up programme CCC and a comparison of its results with NPD experiments

    International Nuclear Information System (INIS)

    Hoejerup, C.F.

    1976-10-01

    A brief description is given of the computer programme CCC, which can be used for rod/rod cluster burn up calculations. A comparison of CCC results with some Canadian measurements on NPD fuel is also included. (author)

  20. Calculational prediction of fuel burn-up for the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Phuoc Lan; Do Quang Binh

    2016-01-01

    In this paper, the method of expanding operators and functions in the neutron diffusion equations as chains of time variable is used for calculation of fuel burn-up of the Dalat nuclear reactors. A computer code, named BURREF, programmed in language Fortran-77 running on IBM PC-AT, has been developed based on this method to predict the fuel burn-up of the Dalat reactor. Some results will be presented here. (author)

  1. MTR fuel element burn-up measurements by the reactivity method

    International Nuclear Information System (INIS)

    Zuniga, A.; Cuya, T.R.; Ravnik, M.

    2003-01-01

    Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

  2. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  5. The burn-up credit physics and the 40. Minerve anniversary; La physique du credit Burn-Up et le 40. anniversaire de Minerve

    Energy Technology Data Exchange (ETDEWEB)

    Santamarina, A [CEA/Cadarache, Departement d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Toubon, H [Cogema, 78 - Velizy Villacoublay (France); Trakas, C [FRAMATOME, 92 - Paris La Defense (France); and others

    2000-03-21

    The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

  6. On the condition of UO{sub 2} nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

    Energy Technology Data Exchange (ETDEWEB)

    Restani, R.; Horvath, M. [Paul Scherrer Institut, CH-5232, Villigen PSI (Switzerland); Goll, W. [AREVA GmbH, P.O. Box 1109, DE-91001 Erlangen (Germany); Bertsch, J.; Gavillet, D.; Hermann, A. [Paul Scherrer Institut, CH-5232, Villigen PSI (Switzerland); Martin, M., E-mail: matthias.martin@psi.ch [Paul Scherrer Institut, CH-5232, Villigen PSI (Switzerland); Walker, C.T. [The Grange, 66 High Street, Swinderby, Lincoln LN6 9LU (United Kingdom)

    2016-12-01

    Post-irradiation examination results are presented for UO{sub 2} fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain. - Highlights: • Gas retention measured by laser ablation induction coupled plasma mass spectrometry. • Thermal release from the high burn structure responsible for high gas release. • At a pellet burn-up of 115 MWd/kgHM the high burn-up structure is still evolving. • The gas pressure in HBS pores is well below the pressure that the fuel can sustain.

  7. Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

    International Nuclear Information System (INIS)

    El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

    2004-01-01

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

  8. Development of materials for higher burn-up rates

    International Nuclear Information System (INIS)

    Ehrlich, K.; Anderko, K.

    1983-01-01

    The results obtained so far concerning both high-nickel austenitic steels (X10CrNiMoTiB10 25) and ferritic-martensitic steels (X18CrMoVNb12 1 corresponds to 1.4914) justify the expectations placed in these materials. Future R and D work will concentrate on high-nickel austenitic steels, in particular on in-pile residence time in the range of maximum cladding mean temperatures (He embrittlement), on confirmation of the fairly good swelling and creep behaviour to be obtained from neutron irradiation experiments as well as on material qualification and technology. As far the development in the field of martensitic steels the major question concerns the influence of neutron irradiation on the transient temperature. It is just as important and interesting to detect and describe the ruling in-pile creep mechanism in the 400 to 600 0 C range. Furthermore, questions on the fabrication and qualification of cladding tube boxes are also preeminent. (orig.) [de

  9. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  10. Burn-up determination of irradiated thoria samples by isotope dilution-thermal ionisation mass spectrometry

    International Nuclear Information System (INIS)

    Aggarwal, S.K.; Jaison, P.G.; Telmore, V.M.; Shah, R.V.; Sant, V.L.; Sasibhushan, K.; Parab, A.R.; Alamelu, D.

    2010-03-01

    Burn-up was determined experimentally using thermal ionization mass spectrometry for two samples from ThO 2 bundles irradiated in KAPS-2. This involved quantitative dissolution of the irradiated fuel samples followed by separation and determination of Th, U and a stable fission product burn-up monitor in the dissolved fuel solution. Stable fission product 148 Nd was used as a burn-up monitor for determining the number of fissions. Isotope Dilution-Thermal Ionisation Mass Spectrometry (ID-TIMS) using natural U, 229 Th and enriched 142 Nd as spikes was employed for the determination of U, Th and Nd, respectively. Atom % fission values of 1.25 ± 0.03 were obtained for both the samples. 232 U content in 233 U determined by alpha spectrometry was about 500 ppm and this was higher by a factor of 5 compared to the theoretically predicted value by ORIGEN-2 code. (author)

  11. Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA

    International Nuclear Information System (INIS)

    Toubon, H.; Riffard, C.; Batifol, M.; Pelletier, S.

    2003-01-01

    For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)

  12. Burn-up function of fuel management code for aqueous homogeneous reactors and its validation

    International Nuclear Information System (INIS)

    Wang Liangzi; Yao Dong; Wang Kan

    2011-01-01

    Fuel Management Code for Aqueous Homogeneous Reactors (FMCAHR) is developed based on the Monte Carlo transport method, to analyze the physics characteristics of aqueous homogeneous reactors. FMCAHR has the ability of doing resonance treatment, searching for critical rod heights, thermal hydraulic parameters calculation, radiolytic-gas bubbles' calculation and bum-up calculation. This paper introduces the theory model and scheme of its burn-up function, and then compares its calculation results with benchmarks and with DRAGON's burn-up results, which confirms its bum-up computing precision and its applicability in the bum-up calculation and analysis for aqueous solution reactors. (authors)

  13. Validation of a continuous-energy Monte Carlo burn-up code MVP-BURN and its application to analysis of post irradiation experiment

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio

    2000-01-01

    In order to confirm the reliability of a continuous-energy Monte Carlo burn-up calculation code MVP-BURN, it was applied to the burn-up benchmark problems for a high conversion LWR lattice and a BWR lattice with burnable poison rods. The results of MVP-BURN have shown good agreements with those of a deterministic code SRAC95 for burn-up changes of infinite neutron multiplication factor, conversion ratio, power distribution, and number densities of major fuel nuclides. Serious propagation of statistical errors along burn-up was not observed even in a highly heterogeneous lattice. MVP-BURN was applied to the analysis of a post irradiation experiment for a sample fuel irradiated up to 34.1 GWd/t, together with SRAC95 and SWAT. It was confirmed that the effect of statistical errors of MVP-BURN on a burned fuel composition was sufficiently small, and it could give a reference solution for other codes. In the analysis, the results of the three codes with JENDL-3.2 agreed with measured values within an error of 10% for most nuclides. However, large underestimation by about 20% was observed for 238 Pu, 242m Am and 244 Cm. It is probable that these discrepancies are a common problem for most current nuclear data files. (author)

  14. FBR type reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  15. FBR type reactors

    International Nuclear Information System (INIS)

    Maemoto, Junko.

    1985-01-01

    Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

  16. Comparison of seismic isolation concepts for FBR

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  18. Reactivity management and burn-up management on JRR-3 silicide-fuel-core

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

    2007-08-01

    On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)

  19. Experimental methods for burn-up determination in nuclear fuels, 1

    International Nuclear Information System (INIS)

    Taddei, J.F. de A.C.; Rodrigues, C.

    1977-01-01

    A method is presented that allows the calculation of the total percentage of atoms having undergone fission ('burn up') in nuclear fuels, from the measurement of absolute amounts of fission product neodymium-148 and of uranium and plutoniun present in the spent fuel, the fission yield of neodymium-148 being known. These measurements are performed through the mass spectrometry- isotope dilution technique [pt

  20. Reactivity effect of spent fuel depending on burn-up history

    International Nuclear Information System (INIS)

    Hayashi, Takafumi; Suyama, Kenya; Nomura, Yasushi

    2001-06-01

    It is well known that a composition of spent fuel depends on various parameter changes throughout a burn-up period. In this study we aimed at the boron concentration and its change, the coolant temperature and its spatial distribution, the specific power, the operation mode, and the duration of inspection, because the effects due to these parameters have not been analyzed in detail. The composition changes of spent fuel were calculated by using the burn-up code SWAT, when the parameters mentioned above varied in the range of actual variations. Moreover, to estimate the reactivity effect caused by the composition changes, the criticality calculations for an infinite array of spent fuel were carried out with computer codes SRAC95 or MVP. In this report the reactivity effects were arranged from the viewpoint of what parameters gave more positive reactivity effect. The results obtained through this study are useful to choose the burn-up calculation model when we take account of the burn-up credit in the spent fuel management. (author)

  1. FBR fuel cycle can be competitive

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  2. Summary: analysis of alternative FBR development strategies

    International Nuclear Information System (INIS)

    Burnham, J.B.

    1981-12-01

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

  3. FBR Plant Engineering Center annual report 2012

    International Nuclear Information System (INIS)

    2013-12-01

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

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

    International Nuclear Information System (INIS)

    Kani, Yoshio

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Ono, Kiyoshi; Shiotani, Hiroki; Hirao, Kazunori

    2003-07-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  8. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  9. Technique for sensitivity analysis of space- and energy-dependent burn-up calculations

    International Nuclear Information System (INIS)

    Williams, M.L.; White, J.R.

    1979-01-01

    A practical method is presented for sensitivity analysis of the very complex, space-energy dependent burn-up equations, in which the neutron and nuclide fields are coupled nonlinearly. The adjoint burn-up equations that are given are in a form which can be directly implemented into multi-dimensional depletion codes, such as VENTURE/BURNER. The data sensitivity coefficients can be used to determine the effect of data uncertainties on time-dependent depletion responses. Initial condition sensitivity coefficients provide a very effective method for computing the change in end of cycle parameters (such as k/sub eff/, fissile inventory, etc.) due to changes in nuclide concentrations at beginning of cycle

  10. The burn-up credit physics and the 40. Minerve anniversary

    International Nuclear Information System (INIS)

    Santamarina, A.; Toubon, H.; Trakas, C.

    2000-01-01

    The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

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

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

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

  12. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  13. Evaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste - 13187

    Energy Technology Data Exchange (ETDEWEB)

    Fast, Ivan; Aksyutina, Yuliya; Tietze-Jaensch, Holger [Product Quality Control Office for Radioactive Waste (PKS) at the Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum Juelich (Germany)

    2013-07-01

    Burn up calculations facilitate a determination of the composition and nuclear inventory of spent nuclear fuel, if operational history is known. In case this information is not available, the total nuclear inventory can be determined by means of destructive or, even on industrial scale, nondestructive measurement methods. For non-destructive measurements however only a few easy-to-measure, so-called key nuclides, are determined due to their characteristic gamma lines or neutron emission. From these measured activities the fuel burn up and cooling time are derived to facilitate the numerical inventory determination of spent fuel elements. Most regulatory bodies require an independent assessment of nuclear waste properties and their documentation. Prominent part of this assessment is a consistency check of inventory declaration. The waste packages often contain wastes from different types of spent fuels of different history and information about the secondary reactor parameters may not be available. In this case the so-called characteristic fuel burn up and cooling time are determined. These values are obtained from a correlations involving key-nuclides with a certain bandwidth, thus with upper and lower limits. The bandwidth is strongly dependent on secondary reactor parameter such as initial enrichment, temperature and density of the fuel and moderator, hence the reactor type, fuel element geometry and plant operation history. The purpose of our investigation is to look into the scaling and correlation limitations, to define and verify the range of validity and to scrutinize the dependencies and propagation of uncertainties that affect the waste inventory declarations and their independent verification. This is accomplished by numerical assessment and simulation of waste production using well accepted codes SCALE 6.0 and 6.1 to simulate the cooling time and burn up of a spent fuel element. The simulations are benchmarked against spent fuel from the real reactor

  14. Optimalisation Of Oxide Burn-Up Enhanced For RSG-Gas Core

    International Nuclear Information System (INIS)

    Tukiran; Sembiring, Tagor Malem

    2000-01-01

    Strategy of fuel management of the RSG-Gas core has been changed from 6/1 to 5/1 pattern so the evaluation of fuel management is necessary to be done. The aim of evaluation is to look for the optimal fuel management so that the fuel can be stayed longer in the core and finally can save cost of operation. Using Batan-EQUIL-2D code did the evaluation of fuel management with 5/1 pattern. The result of evaluation is used to choose which one is more advantage without break the safety margin which is available in the Safety Analysis Report (SAR) firstly, the fuel management was calculated with core excess reactivity of 9,2% criteria. Secondly, fuel burn-up maximum of 56% criteria and the last, fuel burn-up maximum of 64% criteria. From the result of fuel management calculation of the RSG-Gas equilibrium core can be concluded that the optimal RSG-Gas equilibrium core with 5/1 pattern is if the fuel burn-up maximum 64% and the energy in a cycle of operation is 715 MWD. The fuel can be added one more step in the core without break any safety margin. It means that the RSG-Gas equilibrium core can save fuel and cost reduction

  15. Burn up determination of IEAR-1 fuel elements by non destructive gamma ray spectrometry method

    International Nuclear Information System (INIS)

    Soares, A.J.

    1977-01-01

    Measurement of nuclear fuel burn up by non destructive gamma ray spectrometry is discussed, and results of such measurements, made at the Instituto de Energia Atomica (IEA), are given. Specifically, the burn up of an MTR (Material Testing Reactor) fuel element removed from the IEAR-1 swimming pool reactor in 1958 is evaluated from the measured Cs-137 activity, which gives a single 661,6 keV gamma ray. Due to the long decay time of the test element, no other fission decay product activity could be detected. Analysis of measurements, made with a 3'' x 3'' NaI(Tl) detector at 330 distinct points of the element, showed the total burn up to 3.3 +- -+ 0.8 mg. This is in agreement with a calculated value. As the maximum temperature of IEAR-1 fuel elements is of the order of 40 0 C, migration effects of Cs-137 was not considered, this being significant only at fuel temperature in excess of 1000 0 C [pt

  16. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

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

  17. Non destructive burn up determination of IEA-R1 reactor fuel elements by gamma-ray spectrometry using a Ge(Li) detector

    International Nuclear Information System (INIS)

    Madi Filho, T.

    1982-01-01

    A non destructive determination of burn up of low (IEA-14) and high (IEA-80) activity fuel elements used in the IEA-R1 pool reactor was made from the measured distribution of the Cs-137 gamma-ray activity in these elements. For both series of measurements a 73,7 c.c. Ge(Li) detector was used in 'well collimated' geometry. Where as IEA-14, removed from the reactor some 20 years, showed a gamma-ray spectrum essentially due to Cs-137, IEA-80, with a cooling time of 5 years, showed a more complex spectrum due to the greater number of fission products remaining. The S.I out-of-pool assembly was calibrated using Cs-137 and Co-60 point and Ag-110m plane sources. These measurements provided the necessary constants used to calculate fuel burn-up from measured relative activity distributions of fuel elements. Detailed fuel plate transmission measurements made with the Cs-137 source showed the plates to be highly homogeneous. High activity fuel elements were measured in the S.II in-pool assembly in which the detector was locate on the moveable pool bridge and the test element was positioned immediately below the detector 2.17m below the pool surface. Measurements made in the S.II assembly were normalised with respect to the measured activity of the IEA-14 element. The measured burn up of the IEA-14 and IEA-80 elements obtained in this work is 3.22.10 - 3 gms and 24.44gms. These values may be compared with respective values of 2.63.10 - 3 gms and 61.11gms given by 'total reactor energy/flux distribution' calculations. Calculated errors for the U-235 burn up are 7.4% (IEA-14) and 10.1% (IEA-80). A detailed evaluation of the errors associated with both sets of measurements is given. (Author) [pt

  18. Creep fatigue design of FBR components

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Meshii, Toshiyuki; Asayama, Tai

    2001-03-01

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

  20. Challenges in the application of burn-up credit to the criticality safety of the THORP reprocessing plant

    International Nuclear Information System (INIS)

    Mayson, R.T.H.; Gunston, K.J.

    1999-01-01

    Since 1991 BNFL has made a significant investment in the development of the burn-up credit method and the application to its operations. It has recently demonstrated that using this method for the THORP dissolvers, it is possible to justify operating safety with reduced neutron poison concentrations and this has now been submitted to the regulators. The continued challenges the criticality safety community is facing are to show that we are not reducing safety levels because we are using burn-up credit. The burn-up credit method that has been developed can be summarized as follows. It consists of performing reactivity calculations for irradiated fuel using compositions generated by and inventory prediction code, generally in order to determine the limiting burn-up required for that fuel in a particular environment. In addition, it has always been envisaged that a confirmatory measurement of burn-up would be required to be made prior to certain operations such as the sharing of fuel into a dissolver. The burn-up credit method therefore relies upon three key components of inventory prediction, reactivity calculation code and the quantification and verification of burn-up. (J.P.N.)

  1. Design study on metal fuel FBR cores

    International Nuclear Information System (INIS)

    Yokoo, T.; Tanaka, Y.; Ogata, T.

    1991-01-01

    A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

  2. Seismic design for Monju FBR power plant

    International Nuclear Information System (INIS)

    1982-01-01

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

  3. Characterization of alternative FBR development strategies

    International Nuclear Information System (INIS)

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

    1981-08-01

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

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

    International Nuclear Information System (INIS)

    Yoshimi, H.; Hachiya, Y.

    1990-01-01

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

  5. Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

    International Nuclear Information System (INIS)

    Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

    2010-01-01

    Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.

  6. Sensitivity change of rhodium self -powered detectors with burn-up

    International Nuclear Information System (INIS)

    Girgis, R.; Akimov, I.S.; Hamouda, I.

    1976-01-01

    The scope of the present paper is to obtain the calculation formulae to evaluate the rate of sensitivity change of the neutron self-powered detectors with burn-up. A code written in FORTRAN 4 was developed to be operational on the IBM-1130 computer. It has been established in the case of rhodium detectors that neglecting the β-particle absorption in the calculations leads to the underestimation of the detector sensitivity decrease up to 40%. The derived formulae can be used for other self-powered detectors. (author)

  7. Burn up Theoretical Analysis of A Thorium Fuel Rod in Light Water Reactor

    International Nuclear Information System (INIS)

    Gaber, F.A.; Aziz, M.; Elsheikh, B.

    2008-01-01

    A computer model was designed to analyze the burn up and irradiation of both Th-Pu and Th-U fuel rod in a typical light water reactors conditions. MCNP computer model was designed to simulate the fuel rod burnup and evaluate neutron flux and group constants . A system of ordinary differential equations were solved numerically to evaluate the isotopic concentrations for both the two types of fuel using the previous calculated data from MCNP model. The results are analyzed and compared with published data where satisfactory agreement was found

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

    International Nuclear Information System (INIS)

    Higuchi, Masahisa

    1990-01-01

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

  9. Study on small long-life LBE cooled fast reactor with CANDLE burn-up. Part 1. Steady state research

    International Nuclear Information System (INIS)

    Yan, Mingyu; Sekimoto, Hiroshi

    2008-01-01

    Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is used as coolant. From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%. (author)

  10. Burn-Up Calculation of the Fuel Element in RSG-GAS Reactor using Program Package BATAN-FUEL

    International Nuclear Information System (INIS)

    Mochamad Imron; Ariyawan Sunardi

    2012-01-01

    Calculation of burn lip distribution of 2.96 gr U/cc Silicide fuel element at the 78 th reactor cycle using computer code program of BATAN-FUEL has been done. This calculation uses inputs such as generated power, operation time and a core assumption model of 5/1. Using this calculation model burn up for the entire fuel elements at the reactor core are able to be calculated. From the calculation it is obtained that the minimum burn up of 6.82% is RI-50 at the position of A-9, while the maximum burn up of 57.57% is RI 467 at the position of 8-7. Based on the safety criteria as specified in the Safety Analysis Report (SAR) RSG-GAS reactor, the maximum fuel burn up allowed is 59.59%. It then can be concluded that pattern that elements placement at the reactor core are properly and optimally done. (author)

  11. Sintered-to-size FBR fuel

    International Nuclear Information System (INIS)

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

    1984-04-01

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

  12. MODRIB - a zero dimensional code for criticality and burn-up of LWR's

    International Nuclear Information System (INIS)

    Gaafar, M.A.; El-Cherif, A.I.

    1980-01-01

    The computer program MODRIB is a zero-dimensional code for calculating criticality and burn-up of light water reactors (LWR's). It is a version of an Italian code RIBOT-2 with an updated cross-section data library. The nuclear constants of MODRIB-code are calculated with a two group scheme (fast and thermal), where the fast group is an average of three fast groups. The code requires as input data essential extensive reactor parameters such as fuel rod radius, clad thickness, fuel enrichment, lattice pitch, water density and temperature etc. A summary of the physical model description and the input-output procedures are given in this report. Selected results of two sample problems are also given for the purpose of checking the validity and reliability of the code. The first is BWR and the second is PWR. The calculation time for a criticality problem with burn-up is about 8 seconds for the first time step and about 3 seconds for each subsequent time step on the ICL-1906 computer facility. The requirements on the memory size is less than 32 K-word. (author)

  13. Burn-up measurements of spent fuel using gamma spectrometry technique

    International Nuclear Information System (INIS)

    Pereda, C.; Henriquez, C.; Klein, J.; Medel, J.

    2005-01-01

    Burn-up results obtained for HEU (45% of 235 U) fuel assemblies of the RECH-1 Research Reactor using gamma spectrometry technique are presented. The spectra were got from an in-pool facility built in the reactor to be mainly used to measure the burnup of irradiated fuel assemblies with short cooling time, where 95 Zr is being evaluated as possible fission monitor. A program to measure all spent fuel assemblies of the RECH-1 reactor was initiated in the frame of the Regional Project RLA/4/018: 'Management of Spent Fuel from Research Reactors'. The results presented here were obtained from HEU spent fuel assemblies with cooling time greater than 100 days and 137 Cs was used as fission monitor. The efficiency of the in-pool system was determined using a slightly burnt experimental fuel assembly, which has one fuel plate (one of the outer plates) and the rest are dummy plates. An average burn-up of 2.8% of 235 U was previously measured for the experimental fuel assembly utilizing a facility installed in a hot cell and 137 Cs was used as monitor. (author)

  14. The estimation of the control rods absorber burn-up during the VVER-1000 operation

    Energy Technology Data Exchange (ETDEWEB)

    Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2013-09-15

    The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

  15. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  16. A validated methodology for evaluating burn-up credit in spent fuel casks

    International Nuclear Information System (INIS)

    Brady, M.C.; Sanders, T.L.

    1992-01-01

    The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (USDOE) programme to resolve issues related to the implementation of burn-up credit in spent fuel cask design. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burn-up credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor re-start critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias effective multiplication (k eff ). Implementation issues affecting licensing requirements and operational procedures are discussed briefly. (Author)

  17. Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

    International Nuclear Information System (INIS)

    Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

    2006-01-01

    A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

  18. Calculation of fuel burn-up and fuel reloading for the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lan, Nguyen Phuoc; Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Binh, Do Quang [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Calculation of fuel burnup and fuel reloading for the Dalat Nuclear Research Reactor was carried out by using a new programme named HEXA-BURNUP, realized in a PC. The programme is used to calculate the following parameters of the Dalat reactor: a/Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92 fuel elements. The effective multiplication coefficients equal 1 within the error ranges of less than 0.38%. b/ The thermal neutron flux distribution in the reactor. The calculated results agree with the experimental data measured at 11 typical positions. c/The average fuel burn-up for the period from Feb. 1984 to Sep. 1992. The difference between calculation and experiment is only about 1.9%. 10 fuel reloading versions are calculated, from which an optimal version is proposed. (author). 9 refs., 4 figs., 5 tabs.

  19. Burn-up analysis of uranium silicide fuels 20% 235U, in the LFR facility

    International Nuclear Information System (INIS)

    Amor, Ricardo A.; Bouza, Edgardo; Cabrejas, Julian L.; Devida, Claudio A.; Gil, Daniel A.; Stankevicius, Alejandro; Gautier, Eduardo; Garavaglia, Ricardo N.; Lobo, Alfredo

    2003-01-01

    The LFR Facility is a laboratory designed and constructed with a Hot-Cells line, a Globe-Box and a Fume-Hood, all of them suited to work with radioactive materials such as samples of irradiated silicide MTR fuel elements. A series of dissolutions of this material was performed. From the resulting solutions, two fractions were separated by HPLC. One contained U + Pu, and other the fission product Nd. The concentrations of these elements were obtained by isotopic dilution and mass spectrometry (IDMS). It is concluded that this technique is very powerful and accurate when properly applied, and makes the validation of burn-up calculation codes possible. It is worth remarking the Lfr capacity to carry on different Research and Development (R + D) tasks in the Nuclear Fuel Cycle field. (author)

  20. Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

    International Nuclear Information System (INIS)

    Belo, Thiago F.; Fiel, Joao Claudio B.

    2015-01-01

    Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

  1. Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

    Energy Technology Data Exchange (ETDEWEB)

    Belo, Thiago F.; Fiel, Joao Claudio B., E-mail: thiagofbelo@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  3. FBR pellet fabrication - density and dimensional control

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  5. Development of FBR cycle data base system

    International Nuclear Information System (INIS)

    Kubota, Sadae; Ohtaki, Akira; Hirao, Kazuhiro

    2002-06-01

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

  6. Practicalization strategic research of FBR cycle

    International Nuclear Information System (INIS)

    2000-01-01

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

  7. Studies on use of reflector material and its position within FBR core for reducing U{sup 232} content of U produced in ThO{sub 2} radial blankets

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Sujoy, E-mail: sujoy@igcar.gov.in [Core Design Group, IGCAR, Kalpakkam (India); Prasad, Rajeev Ranjan; Bagchi, Subhrojit [Core Design Group, IGCAR, Kalpakkam (India); Mohanakrishnan, P. [MCNS, Manipal University, Manipal (India); Arul, A. John; Puthiyavinayagam, P. [Core Design Group, IGCAR, Kalpakkam (India)

    2015-11-15

    Highlights: • Nuclear data processing for multigroup neutron transport calculation. • Discrete ordinate and Monte Carlo neutron transport. • Breeding of Thorium in Fast Reactor. • Minimization of U{sup 232} in U{sup 233}. • Fuel burn up using Neutron Diffusion. - Abstract: Presence of U{sup 232} in U{sup 233} bred in thorium blanket of fast reactor is a major concern in fuel reprocessing. The former's daughter products being hard gamma emitter and the isotope itself having substantial half life, its presence beyond 10 ppm makes fuel recycle complicated and expensive. In this study possibility of decreasing U{sup 232} production in a typical FBR blanket by means of spectrum modification is examined. SS, depleted B{sub 4}C, SiC, Mo and W regions were introduced between core and radial blanket and evolution of isotopes were studied to arrive at an optimal configuration that satisfies requirements of breeding U{sup 233} and lowering U{sup 232}concentration. SS, B{sub 4}C, SiC, Mo and W are known to be high temperature material with appropriate stability in harsh fast reactor environment. Study has shown that introducing two SS reflector rows can achieve the required low value of U{sup 232}concentration without greatly compromising the U{sup 233}production.

  8. Development of a depletion program for the calculation of the 3D-burn-up dependent power distributions in light water reactors

    International Nuclear Information System (INIS)

    Bennewitz, F.; Mueller, A.; Wagner, M.R.

    1977-11-01

    Based on the nodal collision probability method a multi-dimensional reactor burn-up program MEDIUM has been developed, which is written for 2 neutron energy groups. It is characterized by high computing speed, considerable generality and flexibility, a number of useful program options and good accuracy. The three-dimensional flux calculation model is described, the formulation and method of solution of the nuclear depletion equations and further details of the program structure. The results of a number of comparisons with experimental data and with independent computer programs are presented. (orig.) [de

  9. Development of destructive methods of burn-up determination and their application on WWER type nuclear fuels

    International Nuclear Information System (INIS)

    Hermann, A.; Stephan, H.; Nebel, D.

    1984-03-01

    Results are described of a cooperation between the Central Institute of Nuclear Research Rossendorf and the Radium Institute 'V.G. Chlopin' Leningrad in the field of destructive burn-up determination. Laboratory methods of burn-up determination using the classical monitors 137 Cs, 106 Ru, 148 Nd and isotopes of heavy metals (U, Pu) as well as the usefulness of 90 Sr, stable isotopes of Ru and Mo as monitors are dealt with. The analysis of the fuel components uranium (spectrophotometry, potentiometric titration, mass-spectrometric isotope dilution) and plutonium (spectrophotometry, coulometric titration, mass- and alpha-spectrometric isotope dilution) is fully described. Possibilities of increasing the reproducibility (automatic adjusting of measurement conditions) and the sensibility (ion impuls counting) of mass-spectrometric measurements are proposed and applied to a precise determination of Am and Cm isotopic composition. The methods have been used for burn-up analysis of spent WWER (especially WWER-440) fuel. (author)

  10. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  11. Monte Carlo sampling on technical parameters in criticality and burn-up-calculations

    International Nuclear Information System (INIS)

    Kirsch, M.; Hannstein, V.; Kilger, R.

    2011-01-01

    The increase in computing power over the recent years allows for the introduction of Monte Carlo sampling techniques for sensitivity and uncertainty analyses in criticality safety and burn-up calculations. With these techniques it is possible to assess the influence of a variation of the input parameters within their measured or estimated uncertainties on the final value of a calculation. The probabilistic result of a statistical analysis can thus complement the traditional method of figuring out both the nominal (best estimate) and the bounding case of the neutron multiplication factor (k eff ) in criticality safety analyses, e.g. by calculating the uncertainty of k eff or tolerance limits. Furthermore, the sampling method provides a possibility to derive sensitivity information, i.e. it allows figuring out which of the uncertain input parameters contribute the most to the uncertainty of the system. The application of Monte Carlo sampling methods has become a common practice in both industry and research institutes. Within this approach, two main paths are currently under investigation: the variation of nuclear data used in a calculation and the variation of technical parameters such as manufacturing tolerances. This contribution concentrates on the latter case. The newly developed SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool) is introduced. It defines an interface to the well established GRS tool for sensitivity and uncertainty analyses SUSA, that provides the necessary statistical methods for sampling based analyses. The interfaced codes are programs that are used to simulate aspects of the nuclear fuel cycle, such as the criticality safety analysis sequence CSAS5 of the SCALE code system, developed by Oak Ridge National Laboratories, or the GRS burn-up system OREST. In the following, first the implementation of the SUnCISTT will be presented, then, results of its application in an exemplary evaluation of the neutron

  12. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  13. Establishment of THERPRO Database and Estimation of the Effect of Fuel Burn-up on the Thermal Conductivity of Uranium Dioxide

    International Nuclear Information System (INIS)

    Lee, Hyun Seon

    2005-02-01

    Materials property data are an essential part of major disciplines in many engineering fields. To nuclear engineering, fundamental understanding of thermo-physical chemical mechanical properties of nuclear materials is very important. THERPRO data base that is re-designed and re-constructed through this study is a web-based on-line nuclear materials properties data base. For the future upgrade of the data base contemporary information technologies have been incorporated during the construction. Basically THERPRO data base has a hierarchical structure consisting of several levels: home page, element, compound, property, author, report, and bibliography level. All of data sets in each level are interconnected using network structure and thus every data can be easily retrieved including the bibliographical information by an appropriate query action. As a part of THERPRO DB utilization, the effect of fuel burn-up on the thermal conductivity of irradiated uranium dioxide is analyzed with the data contained in the data base as well as recent data published in the relevant journals. Their data are comparatively studied and the effect is estimated using FRAPCON-3 code with two in-pile data sets, BR-3 111i5 and Oconee rod 15309. The results show that the fuel center line temperature can differ 200 .deg. C∼400 .deg. C from thermal conductivity models depending on burn-up, which can significantly influence high burn-up fuel performance. In conclusion, it is demonstrated through this study that THERPRO data base can be a great utility for nuclear engineers and researchers, if appropriately utilized

  14. Development and validation of ALEPH Monte Carlo burn-up code

    International Nuclear Information System (INIS)

    Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.

    2011-01-01

    The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)

  15. LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

    International Nuclear Information System (INIS)

    Peehs, M.

    1997-01-01

    Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves

  16. LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

    Energy Technology Data Exchange (ETDEWEB)

    Peehs, M [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-12-01

    Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves.

  17. Direct measurement of burn up monitor by Pulsed Laser Deposition (PLD) followed by Isotopic Dilution Mass Spectrometry

    International Nuclear Information System (INIS)

    Sajimol, R.; Manoravi, P.; NaIini, S.; Balasubramanian, R.; Joseph, M.

    2012-01-01

    Burn-up measurement is an important aspect in the assessment of fuel performance especially for experimental nuclear fuels. Conventional mass spectrometric technique offer the best accuracy for determination of burn-up but they suffer from the labour intensive and time consuming chemical separation procedures followed by mass spectrometric analysis. Our laboratory has reported a potential laser mass spectrometric technique with advantages of (i) direct and fast measurement of ion intensities of selected rare earth element and residual heavy element atoms to deduce burn up and (ii) adaptability to remote handling of radioactive samples. Direct quantification of burn up monitor element in fuel in the form of pellet as well as liquid was probed by pulsed laser deposition followed by Isotopic Dilution Mass Spectrometric technique (IDMS). The procedure involving laser ablation of heavy element (namely U and Pu) and fission product (Nd, La etc) from a simulated spent fuel matrix followed by isotopic dilution mass spectrometry using thermal ionization mass spectrometry (TIMS) has been presently attempted to arrive at the rare earth element to heavy element ratio to deduce burn up using the methodology described in our earlier work. The details of IDMS technique has been reviewed by Heumann et al. Accurately weighed amounts of major rare earth fission products such as Nd, La, Ce and Sm in solution form were mixed with known quantity of uranium solution (all the weights are corresponding to their fission yields and the residual heavy element atoms after a given burn up) and mixed together to attain uniformity. The solution is then dried and resulting powder was pelletized and sintered. Subsequently, the pellet was ablated with pulsed laser (8 ns, 532 nm, Nd-YAG) and the plume was deposited on a glass plate. This deposit was dissolved in minimum amount of nitric acid. A known volume of the solution was mixed with spike (for e.g., 150 Nd/ 142 Nd, 233 U/ 238 U in this study

  18. Study on design method for seismically isolated FBR plants

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  19. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

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

  20. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

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

  1. The Swiss contribution to the FBR development

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-05-01

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

  2. The Swiss contribution to the FBR development

    International Nuclear Information System (INIS)

    Hudina, M.

    1981-01-01

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

  3. FBR metallic materials test manual (English version)

    International Nuclear Information System (INIS)

    Odaka, Susumu; Kato, Shoichi; Yoshida, Eiichi

    2003-06-01

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

  4. A state of the Art report on Manufacturing technology of high burn-up fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeong Ho; Nam, Cheol; Baek, Jong Hyuk; Choi, Byung Kwon; Park, Sang Yoon; Lee, Myung Ho; Jeong, Yong Hwan

    1999-09-01

    In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. (author)

  5. Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

    Science.gov (United States)

    Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

    2014-12-01

    Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

  6. High burn-up plutonium isotopic compositions recommended for use in shielding analysis

    International Nuclear Information System (INIS)

    Zimmerman, M.G.

    1977-06-01

    Isotopic compositions for plutonium generated and recycled in LWR's were estimated for use in shielding calculations. The values were obtained by averaging isotopic values from many sources in the literature. These isotopic values should provide the basis for a reasonable prediction of exposure rates from the range of LWR fuel expected in the future. The isotopic compositions given are meant to be used for shielding calculations, and the values are not necessarily applicable to other forms of analysis, such as inventory assessment or criticality safety. 11 tables, 2 figs

  7. FUEL BURN-UP CALCULATION FOR WORKING CORE OF THE RSG-GAS RESEARCH REACTOR AT BATAN SERPONG

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2017-12-01

    Full Text Available The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN and the IAEA (International Atomic Energy Agency. In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U3Si2-Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92, all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.

  8. Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

    2008-01-01

    For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

  9. Development of wire wrapping technology for FBR fuel pin

    International Nuclear Information System (INIS)

    Nogami, Tetsuya; Seki, Nobuo; Sawayama, Takeo; Ishibashi, Takashi

    1991-01-01

    For the FBR fuel assembly, the spacer wire is adopted to maintain the space between fuel pins. The developments have been carried out to achieve automatically wire wrapping with high precision. Based on the fundamental technology developed through the mock-up test operation, Joyo 'MK-I', fuel pin fabrication was started using partially mechanized wire wrapping machine in 1973. In 1978, an automated wire wrapping machine for Joyo 'MK-II' was developed by the adoption of some improvements for the wire inserting system to end plug hole and the precision of wire pitch. On the bases of these experiences, fully automated wire wrapping machine for 'Monju' fuel pin was installed at Plutonium Fuel Production Facility (PFPF) in 1987. (author)

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

    International Nuclear Information System (INIS)

    Miura, Naoki; Nitta, Akito; Take, Kohji

    1988-01-01

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

  11. FBR structural material test facility in flowing sodium environment

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  12. Loop-type FBR reactor

    International Nuclear Information System (INIS)

    Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

    1996-01-01

    The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  14. Calculation of burn-up data for spent LWR-fuels with respect to the design of spent fuel reprocessing plants

    International Nuclear Information System (INIS)

    Gasteiger, R.

    1976-11-01

    The design of spent fuel reprocessing plants makes necessary a detailed knowledge of the composition of the incoming fuels as a function of burn-up. This report gives a broad review on the composition of radionuclides in fuels (fission products, actinides) and structural materials for different burn-up data. (orig.) [de

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

    International Nuclear Information System (INIS)

    Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

    2009-01-01

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

  16. Heat resistant/radiation resistant cable and incore structure test device for FBR type reactor

    International Nuclear Information System (INIS)

    Tanimoto, Hajime; Shiono, Takeo; Sato, Yoshimi; Ito, Kazumi; Sudo, Shigeaki; Saito, Shin-ichi; Mitsui, Hisayasu.

    1995-01-01

    A heat resistant/radiation resistant coaxial cable of the present invention comprises an insulation layer, an outer conductor and a protection cover in this order on an inner conductor, in which the insulation layer comprises thermoplastic polyimide. In the same manner, a heat resistant/radiation resistant power cable has an insulation layer comprising thermoplastic polyimide on a conductor, and is provided with a protection cover comprising braid of alamide fibers at the outer circumference of the insulation layer. An incore structure test device for an FBR type reactor comprises the heat resistant/radiation resistant coaxial cable and/or the power cable. The thermoplastic polyimide can be extrusion molded, and has excellent radiation resistant by the extrusion, as well as has high dielectric withstand voltage, good flexibility and electric characteristics at high temperature. The incore structure test device for the FBR type reactor of the present invention comprising such a cable has excellent reliability and durability. (T.M.)

  17. Burn-up determination of irradiated uranium oxide by means of direct gama spectrometry and by radiochemical method

    International Nuclear Information System (INIS)

    Cunha, I.I.L.; Nastasi, M.J.C.; Lima, F.W.

    1981-09-01

    The burn-up of thermal neutrons irradiated U 3 O 8 (natural uranium) samples has been determined by using both direct gamma spectrometry and radiochemical methods and the results obtained were compared. The fission products 144 Ce, 103 Ru, 106 Ru, 137 Cs and 95 Zr were chosen as burn-up monitors. In order to isolate the radioisotopes chosen as monitors, a radiochemical separation procedure has been established, in which the solvent extraction technique was used to separate cerium, cesium and ruthenium one from the other and all of them from uranium. The separation between zirconium and niobium and of both elements from the other radioisotopes and uranium was accomplished by means of adsorption on a silica-gel column, followed by selective elution of zirconium and of niobium. When use was made of the direct gamma-ray spectrometry method, the radioactivity of each nuclide of interest was measured in presence of all others. For this purpose use was made of gamma-ray spectrometry and of a Ge-Li detector. Comparison of burn-up values obtained by both methods was made by means of Student's 't' test, and this showed that results obtained in each case are statistically equal. (Author) [pt

  18. Evaluation of core distortion in FBR

    International Nuclear Information System (INIS)

    Ikarimoto, I.; Tanaka, M.; Okubo, Y.

    1984-01-01

    The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

  19. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  20. Method for controlling FBR type reactor

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

    1991-01-01

    The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  2. Application of the PSA method to decay heat removal systems in a large scale FBR design

    International Nuclear Information System (INIS)

    Kotake, S.; Satoh, K.; Matsumoto, H.; Sugawara, M.; Sakata, K.; Okabe, A.

    1993-01-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10 -7 /d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  3. Application of the PSA method to decay heat removal systems in a large scale FBR design

    Energy Technology Data Exchange (ETDEWEB)

    Kotake, S; Satoh, K [Japan Atomic Power Company, Otemachi, Chiyoda-ku, Tokyo (Japan); Matsumoto, H; Sugawara, M [Toshiba Corporation (Japan); Sakata, K [Mitsubishi Atomic Power Industries Inc. (Japan); Okabe, A [Hitachi Engineering Co., Ltd. (Japan)

    1993-02-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10{sup -7}/d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  4. The Non-Destructive Determination of Burn-Up by Means of the Prl44 2.18 M Gamma Activity

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Blackadder, W.H.

    1965-05-01

    In recent years, gamma scanning has been used at several establishments for the determination of the burn-up profile along irradiated fuel elements, the 0.75 MeV gamma from Zr-95/Nb-95 being most often employed as the monitored radiation. Difficulties in establishing the geometry and the self-absorption of the gamma activity in the fuel have tended to prevent the application of the method to quantitative burn-up determination, which has usually been carried out by dissolution of selected portions of the fuel followed by conventional fission product separation or by uranium depletion methods. The present paper describes experiments carried out to calibrate a gamma scanner for quantitative measurements by counting the 2.18 MeV gamma activity due to Pr-144, the short-lived daughter of Ce-144 (t 1/2 = 285 days) from selected pellets in several UO 2 fuel specimens. Accurate burn-up values were then determined by dissolution and application of the isotopic dilution method, using stable molybdenum fission products. The elements, which were rotated about their longitudinal axes to minimize asymmetry effects, were viewed by a sodium iodide crystal and a multichannel analyser through a suitable collimator. Correction for attenuation of the gamma activity (much less than for 0.75 MeV) in the fuel elements which were of different diameters (12.6 to 15.04 mm) was made by applying relative attenuation factors and the effective geometry factor of the instrument was determined. In order to check the corrections applied, the counter factor was also calculated, for the 0.75 MeV activity from Zr-95/Nb-95 and in certain cases for the 0.66 MeV activity from Cs-137. The results obtained, demonstrate that at least over the range of diameters and cooling times used the method is suitable for quantitative determinations. Preliminary experiments to explore the possibility of using the high energy gammas (2.35, 2.65 MeV) from Rh-106 as a method for estimating the fraction of fission events

  5. The Non-Destructive Determination of Burn-Up by Means of the Pr{sup l44} 2.18 M Gamma Activity

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Blackadder, W H

    1965-05-15

    In recent years, gamma scanning has been used at several establishments for the determination of the burn-up profile along irradiated fuel elements, the 0.75 MeV gamma from Zr-95/Nb-95 being most often employed as the monitored radiation. Difficulties in establishing the geometry and the self-absorption of the gamma activity in the fuel have tended to prevent the application of the method to quantitative burn-up determination, which has usually been carried out by dissolution of selected portions of the fuel followed by conventional fission product separation or by uranium depletion methods. The present paper describes experiments carried out to calibrate a gamma scanner for quantitative measurements by counting the 2.18 MeV gamma activity due to Pr-144, the short-lived daughter of Ce-144 (t{sub 1/2} = 285 days) from selected pellets in several UO{sub 2} fuel specimens. Accurate burn-up values were then determined by dissolution and application of the isotopic dilution method, using stable molybdenum fission products. The elements, which were rotated about their longitudinal axes to minimize asymmetry effects, were viewed by a sodium iodide crystal and a multichannel analyser through a suitable collimator. Correction for attenuation of the gamma activity (much less than for 0.75 MeV) in the fuel elements which were of different diameters (12.6 to 15.04 mm) was made by applying relative attenuation factors and the effective geometry factor of the instrument was determined. In order to check the corrections applied, the counter factor was also calculated, for the 0.75 MeV activity from Zr-95/Nb-95 and in certain cases for the 0.66 MeV activity from Cs-137. The results obtained, demonstrate that at least over the range of diameters and cooling times used the method is suitable for quantitative determinations. Preliminary experiments to explore the possibility of using the high energy gammas (2.35, 2.65 MeV) from Rh-106 as a method for estimating the fraction of

  6. On-line extraction of the variance caused by burn-up in in-core three-dimensional power distribution

    International Nuclear Information System (INIS)

    Wang Yaqi; Luo Zhengpei; Li Fu; Liu Wenfeng

    2001-01-01

    In most of PWRs, the ex-core ion-chambers are the sole real-time sensors to respond to in-core power and its axial offset. However, the calibration coefficient of the ion-chambers depends on the (3D) power distribution and varies with the burn-up. People expect to know the variance in distribution caused by burn-up directly from the signals of ion-chambers. This expectation is not realized as yet, because an ion-chamber almost only responds to its nearest fuel assemblies. The authors then developed a two-step method for burn-up characteristic extraction: the harmonics synthesis method and harmonics' burn-up grouping. Using the extracted burn-up characteristics, the relationship between the readings of the ex-core ion-chambers and the in-core 3D power distribution is set up. Through the simulation on the heating reactor, the method of burn-up characteristic extraction is verified under engineering conditions. It is possible to on-line extract the variance caused by burn-up in 3D power distribution

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. A comparison study of the 1MeV triton burn-up in JET using the HECTOR and SOCRATE codes

    International Nuclear Information System (INIS)

    Gorini, G.; Kovanen, M.A.

    1988-01-01

    The burn-up of the 1MeV tritons in deuterium plasmas has been measured in JET for various plasma conditions. To interpret these measurements the containment, slowing down and burn-up of fast tritons needs to be modelled with a reasonable accuracy. The numerical code SOCRATE has been written for this specific purpose and a second code, HECTOR, has been adapted to study the triton burn-up problem. In this paper we compare the results from the two codes in order to exclude possible errors in the numerical models, to assess their accuracy and to study the sensitivity of the calculation to various physical effects. (author)

  9. UK regulatory perspective on the application of burn-up credit to the BNFL thorp head end plant

    International Nuclear Information System (INIS)

    Simister, D.N.; Clemson, P.D.

    2003-01-01

    In the UK the Health and Safety Executive, which incorporates the Nuclear Installations Inspectorate (NII), is responsible for regulation of safety on nuclear sites. This paper reports progress made in the application and development of a UK regulatory position for assessing licensee's plant safety caes which invoke the use of Burn-up Credit for criticality applications. The NII's principles and strategy for the assessment of this technical area have been developed over a period of time following expressions of interest from UK industry and subsequent involvement in the international collaborations and debate in this area. This experience has now been applied to the first main plant safety case application claiming Burn-up Credit. This case covers the BNFL Thermal Oxide Reprocessing Plant (THORP) dissolver at Sellafield, where dissolved gadolinium neutron poison is used as a criticality control. The case argues for a reduction in gadolinium content by taking credit for the burn-up of input fuel. The UK regulatory process, assessment principles and criteria are briefly outlined, showing the regulatory framework used to review the case. These issues include the fundamental requirement in UK Health and Safety law to demonstrate that risks have been reduced to as low as reasonably practicable (ALARP), the impact on safety margins, compliance and operability procedures, and the need for continuing review. Novel features of methodology, using a ''Residual Enrichment'' and ''Domain Boundary'' approach, were considered and accepted. The underlying validation, both of criticality methodology and isotopic determination, was also reviewed. Compliance was seen to rely heavily on local in-situ measurements of spent fuel used to determine ''Residual Enrichment'' and other parameters, requiring review of the development and basis of the correlations used to underpin the measurement process. Overall, it was concluded that the case as presented was adequate. Gadolinium reduction

  10. JOYO MK-III performance test. Criticality test, excess reactivity measurement and burn-up coefficient measurement

    International Nuclear Information System (INIS)

    Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki

    2005-03-01

    The MK-III performance test began in June 2003 to fully characterize the upgraded core and heat transfer system of the experimental fast reactor JOYO. This paper describes the results of the approach to criticality, the excess reactivity evaluation and the burn-up coefficient measurement. In the approach to criticality test, the MK-III core achieved initial criticality at the control rod bank position of 412.8 mm on 14:03 July 2nd, 2003. Because the replacement of the outer two rows of reflector subassemblies with shielding subassemblies reduced the source range monitor signals by a factor of 3 at the same reactor power compared with those in the MK-II core, we measured the change of the monitor's response and determined the count rate 2x10 4 cps.' as an appropriate value judging the zero power criticality. In the excess reactivity evaluation, the zero power excess reactivity at 250degC was 2.99±0.10%Δk/kk' based on the measured critical rod bank position and the measured control rod worths. The predicted value by the JOYO core management code system HESTIA was 3.13±0.16%Δk/kk', showing good agreement with the measured value. The measured excess reactivity was within the safety requirement limit. In the burn-up coefficient measurement, the excess reactivity change versus the reactor burn-up was evaluated. The measurement method adopted was to measure the control rod positions during the rated power operation. A value of -2.12x10 -4 Δk/kk'/MWd was obtained as a measured burn-up coefficient. The value calculated by HESTIA was -2.12x10 -4 Δk/kk'/MWd, and it agreed well with the measured value. All technical safety requirements for MK-III core were satisfied and the calculation accuracy of the core management code system HESTIA was confirmed. (author)

  11. Nuclear fuel and/or fertile material element suitable for non-destructive determination of burn-up

    International Nuclear Information System (INIS)

    Muench, E.

    1976-01-01

    The invention refers to a nuclear fuel and/or fertile material element suitable for non-destructive burn-up analysis, where an isotope or a mixture of isotopes capable of being activated is provided for measuring the intensity of radiation emitted from radioactive nuclides, especially the intensity of gamma rays. The half-life of radioactive decay of the isotope or the mixture mentioned above after being activated is sufficiently large compared with the irradiation of the fuel and/or fertile material element in the nuclear reactor. (orig.) [de

  12. Burn-up calculations for a thorium HTR with one and with two types of fuel particle

    Energy Technology Data Exchange (ETDEWEB)

    Griggs, C. F.

    1975-06-15

    Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.

  13. Verification to the RSG-GAS fuel discharge burn-up using SRAC2006 module of COREBN/HIST

    International Nuclear Information System (INIS)

    J-Susilo; T-M-Sembiring; G-R-Sunaryo; M-Imron

    2018-01-01

    For 30 years operation, some of the modifications to the RSG GAS core has been done, that are changes included the type of fuel from U 3 O 8 -Al to U 3 Si 2 -Al with the same density 2.96 gU/cc, the loading pattern of standard fuel elements/fuel control elements from 6/1 & 6/2 to 5/1 pattern, and in core fuel management calculation tool has been change from IAFUEL to BATAN-FUEL. To obtain an extension of the operating license for the next 10 years, the RSG-GAS Periodic Safety Assessment Document is need to prepared. According to the Regulatory Body Chairman Regulation No. 2 2015, RSG-GAS safety assessment should be done independently. As part of this assessment the fuel discharge burn-up must be estimated. In this research, to ensure that the misposition of fuel element in the core has not occurred, the investigation to the document operating report related the fuel placement has been done. Therefore, by using 78 th to 93 rd operation data, verify of the fuel discharge burn-up of the RSG-GAS has been performed by using SRAC2006 module of COREBN/HIST. In addition, the results of these calculations are also made comparative with the operating report data that is calculated by using BATAN-FUEL. Maximum fuel discharge burn-up (57.73 % of U-235) was verified still under permissible value determined by the regulatory body (<60 % of U-235). Maximum differences value between two computer codes was about 2.12 % of U-235 (3.80 %) that is fuel at the B-7 position. Fuel discharge burn-up of RSG-GAS showed almost the same value for each the operation cycle, range of 1.52 % of U-235. So it can be concluded that the RSG-GAS core operation over the last ten years was in good fuel management performance, in accordance with the design. BATAN-FUEL has been conformed well enough with COREBN/HIST. (author)

  14. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  15. Development of FBR piping bellows joint

    International Nuclear Information System (INIS)

    Tsukimori, Kazuyuki; Iwata, Koji

    1991-01-01

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

  16. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  18. Heavy liquid metal cooled FBR. Results 2001

    International Nuclear Information System (INIS)

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

    2003-08-01

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

  19. Burn-up determinations and dimensional measurements of TRIGA-HEU fuel elements from the 14 MW steady-state core

    International Nuclear Information System (INIS)

    Toma, C.; Alexa, Al.; Craciunescu, T.; Pirvan, M.; Dobrin, R.

    2008-01-01

    In this paper there are presented the results of nondestructive examination in Post Irradiation Examination Laboratory for twenty five fuel rods selected from 14 MW steady state core. Gamma scanning and dimensional measurements were carried out in order to determine burn-up and diametric deflection of the fuel rods. Also, some comparisons with SSR Safety Report estimations for the maximum burn-up pin were made. (authors)

  20. Measuring device for the distribution of burn-up degree in fuel assembly irradiated in nuclear reactor

    International Nuclear Information System (INIS)

    Kumanomido, Hironori

    1989-01-01

    The object of the invention is to measure the distribution of burn-up degree, of fuel assemblies irradiated in a nuclear reactor in a short time and exactly. That is, the device comprises a device main body having substantially the same length as that for the axial length of a fuel assembly and a detector container disposed axially slidably to the main body. A plurality of radiation detectors are arranged at an equi-axial pitch and contained in the container. The container is caused to slide at a pitch equal to the equi-axial distance of the detectors. In the device having thus been constituted, measurement is conducted at least for twice at an axial position on the side of a fuel assembly irradiated in the nuclear reactor and a position caused to slide therefrom by one pitch. Based on the result, the sensitivities between each of the detectors are compared and the relative sensitivity of the radiation detectors is calibrated. Accordingly, the sensitivity between each of the detectors can be calibrated rapidly and easily. As a result, the distribution of the burn-up degree, etc of irradiated fuel assembly can be measured exactly. (K.M.)

  1. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    International Nuclear Information System (INIS)

    Toubon, H.; Guillou, E.; Cousinou, P.; Barbry, F.; Grouiller, J.P.; Bignan, G.

    2001-01-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis

  2. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.

  3. Impact of neutron thermal scattering laws on the burn-up analysis of supercritical LWR's fuel assemblies

    International Nuclear Information System (INIS)

    Conti, Andrea

    2011-10-01

    is called the ''free gas approximation''. It is the goal of this work to make an estimate of the criticality calculations' inaccuracy due to the inadequate employed physical model and to determine which one of the available models can be the best replacement. The accuracy of criticality calculations referring to the HPLWR is a problem that had already been raised by Waata in 2006. In her Ph.D. thesis Waata reports having carried out MCNP runs referring to an HPLWR fuel element employing the free gas approximation. In her thesis Waata explicitly sifts through the factors that can affect her MCNP runs' accuracy, but leaves the inappropriate thermal treatment completely out. In this work, the inaccuracy of the criticality calculations has been investigated carrying out sets of similar burn-up calculations differing from each other only in the applied thermal cross section sets. The widest discrepancies were detected between the results obtained applying the free gas model and those obtained applying the molecular models. This, in conjunction with the fact that the free gas model does not even keep in count the molecular structure of H 2 O suggest to discard it and to focus the investigation on the vapour and liquid models. Dr. J. Marti, from the Universitat Politecnica de Catalunya, Barcelona, Spain registered the generalized frequency distributions obtained from the molecular dynamics simulations of 216 molecules of H 2 O in 10 simulated supercritical states and published in an article (1999) the frequencies of the three characteristic distribution peaks for each simulated state, in numerical format. A confrontation with the corresponding peaks from Bernnat's available frequency distributions for liquid water and vapour revealed the peaks of the latter to be closest to the supercritical water ones in nearly all cases. Hence the inference that thermal cross section sets for vapour are for the time being the best replacement for the missing thermal cross section sets for

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

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki

    1999-04-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Kubota, Sadae; Ohtaki, Akira; Hirao, Kazuhiro

    2003-05-01

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

  7. Development of weld plugging for steam generator tubes of FBR

    International Nuclear Information System (INIS)

    Shimoyama, T.; Matsuyama, T.; Matsumoto, O.; Nagura, Y.; Nakamura, H.; Tohguchi, Y.; Kurokawa, M.; Fukada, T.

    2002-01-01

    This study was undertaken to develop a method of weld plugging of the heat-exchanger tubes of steam generator of Prototype FBR 'MONJU' in case these tubes are damaged for some reason. We studied mainly the shape of plug, welding procedure and effect of postweld heat treatment (PWHT). Evaporator tube sheet, tube and plug are made of 2-1/4Cr-1Mo steel and usually preheating and PWHT will be required for welding of this steel. The results of this study is as follows. 1) Plug was designed to make butt joint welding with grooved tube sheet around the tube hole to satisfy the requirements of plug designing, stress analysis, and good weldability. 2) TIG welding process was selected and certified its good weldability and good performance. 3) PWHT can be done by using high frequency induction heating method locally and also designing the plug to weld joint with tube sheet which was grooved around the tube hole. 4) Mock up test was done and it was certified that this plugging procedure has good weldability and good performance ability for Non Destructive Inspection. (author)

  8. Dissolution of mixed oxide spent fuel from FBR

    International Nuclear Information System (INIS)

    Sanyoshi, H.; Nishina, H.; Toyota, O.; Yamamoto, R.; Nemoto, S.; Okamoto, F.; Togashi, A.; Kawata, T.; Hayashi, S.

    1991-01-01

    At the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Chemical Processing Facility (CPF) has been continuing operation since 1982 for laboratory scale hot experiments on reprocessing of FBR mixed oxide fuel. As a part of these experiments, dissolution experiments have been performed to define the key parameters affecting dissolution rates such as concentration of nitric acid, temperature and burnup and also to confirm the amount of insoluble residue. The dissolution rate of the irradiated fuel was determined to be in proportion to the 1.7 power of the nitric acid concentration. The activation energy determined from the experiments varied from 6 to 11 kcal/mol depending on the method of dissolution. The dissolution rate decreased as the fuel burnup increased in low nitric acid media below 5 mol/l. However, it was found that the effect of the burnup became negligible in a high concentration of nitric acid media. The amount of insoluble residue and its constituents were evaluated by changing the dissolution condition. (author)

  9. Research on CDA for advanced fuel FBR

    International Nuclear Information System (INIS)

    Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime.

    1997-03-01

    For the purpose of evaluating possibility of the re-criticality of a metallic fueled reactor, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'Research on CDA for advanced fuel FBR'. The results of this year are the following. The accident initiator considered is a loss-of-flow accident with ATWS. The LOF analysis was performed for the metallic fueled 600 MWe homogeneous two region reactors, both for a metallic fuel only and for a metallic fuel core with ZrH pin. The SAS3D CDA initiation phase analysis code was used to investigate the re-criticality potential at the severe accident. The change mainly in the constants was necessary to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. LOF with flow decay half time of t 1/2 =0.5(s) (all blackout case) and 5.5(s) (ordinary LOF case) were analyzed. Independent of the conditions of the analysis, the results show all the cases could avoid to become prompt-critical. Depending on the analysis condition, it becomes necessary to transfer to the transient phase, it is also shown there is a possibility to avoid re-criticality due to the motion of molten fuel both for the metallic fuel and for the metallic fuel with ZrH moderator. However, because of the constants used for the material property the results might overestimate the fuel motion. It is shown that the moderator is effective to terminate the accident at an early stage. The behavior of metallic fueled reactors at CDA was analyzed with SAS3D code by modifying the constants of material properties to be applied to the reactor. It is shown that a metallic fueled reactor has a possibility to avoid re-criticality at CDA. (J.P.N.)

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

    Science.gov (United States)

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

    2000-01-01

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

  11. Nuclear data needs for the analysis of generation and burn-up of actinide isotopes in nuclear reactors

    International Nuclear Information System (INIS)

    Kuesters, H.

    1980-04-01

    A reliable prediction of the in-pile and out-of-pile physics characteristics of nuclear fuel is one of the objectives of present-day reactor physics. The paper describes the main production paths of important actinides for light water and fast breeder reactors. The accuracy of recent nuclear data is examined by comparisons of theoretical predictions with the results from post-irradiation analysis of nuclear fuel from power reactors, and partly with results obtained in zero-power facilities. A world-wide comparison of nuclear data to be used in large fast power reactor burn-up and long term considerations is presented. The needs for further improvement of nuclear data are discussed. (orig.) [de

  12. Estimating NIRR-1 burn-up and core life time expectancy using the codes WIMS and CITATION

    Science.gov (United States)

    Yahaya, B.; Ahmed, Y. A.; Balogun, G. I.; Agbo, S. A.

    The Nigeria Research Reactor-1 (NIRR-1) is a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria. The reactor went critical with initial core excess reactivity of 3.77 mk. The NIRR-1 cold excess reactivity measured at the time of commissioning was determined to be 4.97 mk, which is more than the licensed range of 3.5-4 mk. Hence some cadmium poison worth -1.2 mk was inserted into one of the inner irradiation sites which act as reactivity regulating device in order to reduce the core excess reactivity to 3.77 mk, which is within recommended licensed range of 3.5 mk and 4.0 mk. In this present study, the burn-up calculations of the NIRR-1 fuel and the estimation of the core life time expectancy after 10 years (the reactor core expected cycle) have been conducted using the codes WIMS and CITATION. The burn-up analyses carried out indicated that the excess reactivity of NIRR-1 follows a linear decreasing trend having 216 Effective Full Power Days (EFPD) operations. The reactivity worth of top beryllium shim data plates was calculated to be 19.072 mk. The result of depletion analysis for NIRR-1 core shows that (7.9947 ± 0.0008) g of U-235 was consumed for the period of 12 years of operating time. The production of the build-up of Pu-239 was found to be (0.0347 ± 0.0043) g. The core life time estimated in this research was found to be 30.33 years. This is in good agreement with the literature

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

    International Nuclear Information System (INIS)

    Kawasaki, Hirotsugu

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. New generation of CASTOR registered casks for high enriched, high burn-up fuel from German NPP

    International Nuclear Information System (INIS)

    Gartz, R.; Kuehne, B.; Diersch, R.

    2004-01-01

    Requirements for new cask designs for transport and long-term dry storage of spent fuel assemblies (FA) from LWR-reactors are based on both increased source terms of the LWR FA including MOX FA, as well as the demand for economical optimisation of decommissioning costs by increased cask capacities. For this, cask development is the challenge to create and establish cask designs that can accommodate more FA with higher source terms, each under fixed boundary conditions (i.e. transport requirements and limitations of the power plants as crane loads and/or fixed maximum dimensions). This task has been elaborated by working simultaneously on different development actions each focussed to improve the cask performance. In the following a brief summary will be presented to give an overview which developments and investigations have been and are still will be performed for development and safety analyses of the new CASTOR registered -designs under the main subjects: material investigation and qualification, component tests and verifications, detailed design analysis and not at least design verification

  16. Simulation of the neutron-physical properties of the classical UO2 fuel and of MOX fuel during the burn-up by Transuranus

    International Nuclear Information System (INIS)

    Breza, J. jr.; Necas, V.; Daoeilek, P.

    2005-01-01

    The classical nuclear fuel UO 2 is well known for VVER reactors. Nevertheless, in the near future it will be possible to replace this fuel by novel, advanced kinds of fuel, for instance MOX, inert matrices fuel, etc., that will allow to increase the level of burn-up and minimize the amount of hazardous waste. The code Transuranus [2], designed at ITU Karlsruhe, is intended for thermal and mechanical analyses of fuel elements in nuclear reactors. We have utilized the code Transuranus to simulate the neutron-physical properties of the classical UO 2 fuel and of MOX fuel during the burn-up to a level of 40 MWd/kgHM. We compare obtained results of uranium and plutonium nuclides concentrations, their changes during burn-up, with results obtained by code HELIOS [3], which is well-validated code for this kind of applications. We performed calculations of fission gasses concentrations, namely xenon and krypton. (author)

  17. Calculations of fuel burn up and radionuclide inventories in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION codes

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-01-01

    The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor as a function of the reactor operating time for 10, 20, and 30 k W operating power levels. The uranium burn up rate and burn up percentage, the amounts of the plutonium isotopes, the concentrations and radioactivities of the fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well. The CITATION code is used to calculate the changes in the effective multiplication factor of the reactor.(author)

  18. Influence of fuel element burn-up on the power peaking factor in PWR; Vpliv zgorelosti gorivnega elementa na konicne faktorje moci v tlacnovodnem reaktorju

    Energy Technology Data Exchange (ETDEWEB)

    Ravnik, M; Mele, I [Institut ' Jozef Stefan' , Ljubljana (Yugoslavia); Falkowski, J [Institut energii atomowel, Swierk (Poland)

    1988-07-01

    Influence of fuel element burn-up distribution on radial power peaking factors is presented for Krsko NPP. The effect is strong for elements loaded in the periphery of the core with large power gradients. Neglecting the burn-up distributions inside fuel elements leads to {+-} 5% error on power peaking factor of the same element and {+-} 2% at other locations in the core. Influence on k is observed due to perturbed leakage from the core and due to redistribution of the importance function of the core. (author)

  19. Burn up physics

    International Nuclear Information System (INIS)

    Tretiakoff, O.

    1964-01-01

    The present communication is devoted to a body of theoretical and experimental work carried out at the C.E.A. with the aim of adding to the current knowledge on the evolution of the reactivity (during fuel irradiation) in natural or slightly enriched Uranium reactors. The difficulties of performing direct experiments on large amounts of irradiated fuels are reviewed - especially in operating power reactors - and the necessity is underlined for fundamental research in two directions: on one hand, the change in the composition of the fuels (chains of heavy nuclei, fission products), and on the other hand the effect of changes in composition on the neutron balance. Before presenting three types of experiments which have been carried out, the importance of the problems associated with the neutron spectra is stressed and the practical methods used for the calculations are briefly described. The systematic irradiation of several types of fuel, followed by their chemical and isotopic analysis has been going on for several years. An outline of the experimental programme is given with a description of the methods employed: α, β, γ chain for the preparation of samples determination of the plutonium content by coulometry and double isotopic dilution, separation of Boron used in some cases for the measurement of integrated neutron densities. The interpretation of the measurements is discussed with some examples. A second and more recent series of experiments deals with the investigation of lattices, using synthetic fuels (Uranium-Plutonium alloys) as compared to slightly depleted or enriched Uranium Various experiments are considered on heavy water and on cold graphite, then on graphite heated up to 500 C Some results already obtained are listed. These experiments, requiring nearly a metric ton of each type of fuel cannot be pursued in a systematic manner. This is why is developed since several years a method of differential measurement by oscillation, which requires samples of the order of several kilograms only. The relationships between these measurements and the investigations of lattices are discussed, and an outline is given of the way of carrying out the systematic study of fuels of various compositions. The method has been successfully applied to the systematic study of irradiated fuels (analysed independently by the methods mentioned above) thus giving the possibility of measuring in situ the absorption of fission products. (author) [fr

  20. A practical approach to burn-up credit use in package design approval for PWR uranium oxide spent fuel assemblies

    International Nuclear Information System (INIS)

    Kroger, H.; Reiche, I.

    2009-01-01

    TN International has applied for a license for the TN 24 E transport and storage cask with the German competent authority using a new Burn-up Credit (BUC) approach for PWR uranium oxide fuel assemblies based on actinides and six selected fission products. In order to enable the use of BUC for fission products, various experimental data have to be provided for the two important aspects of the criticality calculation. Firstly, post-irradiation examination (PIE) experiments for the verification of the calculated fission product concentrations have to be provided for each selected fission product. These data are then used to validate the depletion calculations. Secondly, experimental data for the criticality calculations in the form of critical benchmark experiments have to be provided. The submitted data will be investigated for their applicability to the TN 24 E transport and storage cask. Since the application is limited to six fission products only, the conservatism of the BUC approach can be further justified, as the reduction in reactivity from the remaining fission products (about 190) is not taken credit for. (authors)

  1. Pu Denaturing by Transmutation of MA in FBR Multi-cycle

    Energy Technology Data Exchange (ETDEWEB)

    Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi [Tokyo Institute of Technology, 2-12-1-N1-1 Ookayama, Meguro-ku, Tokyo, 1528550 (Japan)

    2009-06-15

    Pu accumulation and its recycling is important in the term of energy resources, however one of the most sensitive issues is non-proliferation in the future fuel cycle based on fast breeder reactor (FBR). The present paper utilizes Protected Pu Production (P{sup 3}) concept for the production of {sup 238}Pu and {sup 242}Pu by Minor Actinides (MA) transmutation to enhance the proliferation resistance of Pu in the fuel. Increase in the {sup 238}Pu and {sup 242}Pu isotopic fraction creates a high rate of internal heat generation by alpha decay (DH) and/or a high neutron source of spontaneous fission (SFN) in Pu that would be encountered during manufacturing and maintaining of nuclear explosive device. The feasibility of denaturing of Pu by MA transmutation in medium size FBR has been studied from the viewpoint of even-mass number Pu accumulation during multi-cycle of Pu and MA. The proliferation resistance property of Pu is also evaluated based on the specific decay heat and spontaneous fission neutron, compared with the reference criteria. In present paper, the P{sup 3} technology based on multi-recycled Pu and MA is compared with the conventional technology based on multi-recycled Pu only. The detail of mass balance behavior is, however, beyond the scope of the present paper. (authors)

  2. Development of computer code SIMPSEX for simulation of FBR fuel reprocessing flowsheets: II. additional benchmarking results

    International Nuclear Information System (INIS)

    Shekhar Kumar; Koganti, S.B.

    2003-07-01

    Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)

  3. Method of detecting coolant leakages from the pipeways in FBR type reactors

    International Nuclear Information System (INIS)

    Sonoda, Yukio; Tamaoki, Tetsuo

    1986-01-01

    Purpose: To detect coolant leakage in the incore pipeways of loop type FBR type reactors in the initial stage at high sensitivity. Constitution: Temperature of the coolants sealed between incore pipeways and the buffle surrounding them is measured by thermocouples and coolant leakage is detected due to fluctuating components. A well-insertion type in which electrode is sealed with argon is used as the thermo-couples. Signals from the thermocouples are once amplified, removed with DC components and then only the fluctuating components are outputted. The fluctuating components are digitalized, passed through an adaptive digital filter and the RMS value as the difference between the output signal and the thermocouple signal is calculated. The calculated value is compared with a threshold value in a comparative calculator. If it exceeds the threshold value, it is judged as abnormal to display an alarm on an alarm display. In this way, the coolant leakage for the pipeways in the FBR type reactor can be detected on real time and at high sensitivity. (Kamimura, M.)

  4. Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

    Science.gov (United States)

    Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

    2010-10-01

    The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

  5. Changes of the inventory of radioactive materials in reactor fuel from uranium in changing to higher burn-up and determining the important effects of this

    International Nuclear Information System (INIS)

    Kirchner, G.; Schaefer, R.

    1985-01-01

    The knowledge of the nuclide composition during and after use in the reactor is an essential, in order to be able to determine the effects associated with the operation of nuclear plants. The missing reliable data on the inventory of radioactive materials resulting from the expected change to higher burn-ups of uranium fuels in West Germany are calculated. The reliability of the program system used for this, which permits a one-dimensional account taken of the fuel rod cell and measurement of the changes of specific sets of nuclear data depending on burn-up, is confirmed by the comparison with experimentally found concentrations of important nuclides in fuel samples at Obrigheim nuclear power station. Realistic conditions of use are defined for a range of burn-up of 33 GWd/t to 55 GWd/t and the effects of changes of the number of cycles and the use of types of fuel elements being developed on the composition of the inventory are determined. The plutonium compositions during use in the reactor are given and are tabulated with the inventory for decay times up to 30 years. Effects during change to higher burn-ups are examined and discussed for the maximum inventories during use of fuel and for heat generation during final storage. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

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

  7. Correlations among FBR core characteristics for various fuel compositions

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  8. The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

    International Nuclear Information System (INIS)

    Kilger, R.; Gmal, B.; Moser, E.F.

    2008-01-01

    Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

  9. Development of simulation code for FBR spent fuel dissolution with rotary drum type continuous dissolver

    International Nuclear Information System (INIS)

    Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi

    2011-01-01

    Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO 2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (∼500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified. (author)

  10. Summary of the OECD Halden Reactor Project Programme on high burn-up fuel performance relevant for BWRs

    International Nuclear Information System (INIS)

    McGrath, M.A.

    1998-01-01

    The basis for the Halden Reactor Project Programme is presented together with an overview of the content of the programme for the time period 1997-1999. The concept of using both separate effects studies, to determine particular fuel properties, and integral rod behaviour studies of commercial fuel is explained. Each of the items in the programme relevant for BWRs are introduced, with most being discussed in further detail. (author)

  11. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    International Nuclear Information System (INIS)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V.; Boulyga, S.F.; Becker, J.S.

    2005-01-01

    An analytical method is described for the estimation of uranium concentrations, of 235 U/ 238 U and 236 U/ 238 U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10 -9 g/g to 2.0 x 10 -6 g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing 235 U/ 238 U and 236 U/ 238 U isotope ratios and the average value amounted to 9.4±0.3 MWd/(kg U). (orig.)

  12. Preferential removal of Sm by evaporation from Nd-Sm mixture and its application in direct burn-up determination of spent nuclear fuel

    International Nuclear Information System (INIS)

    Sajimol, R.; Bera, S.; Nalini, S.; Sivaraman, N.; Joseph, M.; Kumar, T.

    2016-01-01

    Rate of evaporation of Sm and Nd from their mixture was studied based on their ion intensities using thermal ionization mass spectrometry. Because of the comparatively larger evaporation rate of Sm, it was found possible to get the isotopic composition of Nd (fission product monitor) free from isobaric interference of Sm isotopes. The decrease in ion intensity of Sm was studied as a function of time and filament temperature. Based on this study, an easy and time effective method for the determination of burn-up of spent nuclear fuel was examined and the results are compared with that obtained by the conventional method. Typical burn-up value obtained for a pressurized heavy water reactor fuel dissolver solution using the direct method by preferential evaporation of Sm is: 0.84 at.%, whereas the one obtained by the use of conventional method is 0.82 at.%. In both the cases, Nd was employed as the fission product monitor. (author)

  13. Modification of UO2 grain re-crystallization temperature in function of burn-up as a base for Vitanza experimental curve reconstruction

    International Nuclear Information System (INIS)

    Szuta, M.; Dąbrowski, L.

    2013-01-01

    Crossing the experimental critical fuel temperature dependent on burn-up, an onset of fission gas burst release is observed. This observed phenomena can be explained by assumption that the fission gas immobilization in the uranium dioxide irradiated to a fluency of greater than 10 19 fissions/cm 3 is mainly due to radiation induced chemical activity. Application of the “ab initio” method show that the bond energy of Xenon and Krypton is equal to –1.23 eV, and –3.42 eV respectively. Assuming further that the gas chemically bound can be released mainly in the process of re-crystallization and modifying the differential equation of Ainscough of grain growth by including the burn-up dependence and the experimental data of limiting grain size in function of the fuel temperature for the un-irradiated and irradiated fuel we can re-construct the experimental curve of Vitanza. (authors)

  14. Energy saving options by means of addition of burned-up biomass materials in the ceramics industry; Energiebesparingsmogelijkheden door toevoeging van biomassa-uitbrandstoffen in de keramische industrie

    Energy Technology Data Exchange (ETDEWEB)

    Walda, E.

    2013-06-01

    In 2011/2012 is an exploratory study has been executed on the availability of biomass and the potential applicability in the building ceramics industry. The study consisted of (1) a literature and desk study, in which an overview is made of available and ceramic applicable (renewable) burned-up materials, and (2), laboratory tests in which ultimately potentially applicable burned-up material (sawdust) is examined for its coarse ceramic applicability. In this article the results of the two-pronged research are presented [Dutch] In 2011/2012 is een orienterend onderzoek uitgevoerd naar de beschikbaarheid van biomassa en de mogelijke toepasbaarheid in de bouwkeramische industrie. Het onderzoek bestond uit (1) een literatuur- en deskstudie, waarbij een overzicht is gemaakt van verkrijgbare en keramisch toe te passen (hernieuwbare) uitbrandstoffen, en (2) een laboratoriumonderzoek, waarbij uiteindelijk een potentieel toepasbare uitbrandstof (zaagsel) is onderzocht op zijn grofkeramische toepasbaarheid. In dit artikel worden de resultaten van het tweeledige onderzoek gepresenteerd.

  15. Investigation of the burn-up behavior of boron poison rods, placed in a fuel assembly of a pressurized water reactor

    International Nuclear Information System (INIS)

    Arnold, C.; Lutz, D.C.

    1979-09-01

    The excess reactivity of a pressurized water reactor is compensated by boron, disolved in the moderator. In addition during the first cycle boron poison rods are placed in fuel assemblies without control rods. The burn-up behavior of a poison rod in a Biblis B fuel assembly is analysed in the present paper. Multigroup spectrum calculations were performed. The influence of critical boron concentration depending from burn-up, the changes of fuel concentration and the concentration of burnable poison were taken into consideration. Furthermore the built-up of rapidly saturating fisson products 135 Xe and 149 Sm was considered. The interaction of these effects are discussed. Spatial influences are emphasized most. Finally two group cross sections were calculated. The results are compared with calculations for a fuel assembly of the same type without burnable poison rods. (orig.) [de

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

    International Nuclear Information System (INIS)

    Ishikawa, M.

    2000-01-01

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

  17. Direct Measurement of Initial Enrichment, Burn-up and Cooling Time of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    International Nuclear Information System (INIS)

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-01-01

    An outline of this presentation of what a Differential Die-Away (DDA) instrument can do are: (1) Principle of operation of DDA instrument; (2) Determination of initial enrichment (IE) (σ DDA response increases (die-away time is longer) with increasing fissile content; and (2) Spent fuel => DDA response decreases (die-away time is shorter) with higher burn-up (i.e. more neutron absorbers present).

  18. Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  19. Determination of the burn-up in fuels of the MTR type by means of gamma spectroscopy with crystal of INa(Tl)

    International Nuclear Information System (INIS)

    Kestelman, A.J.

    1988-01-01

    One of the responsibilities of the Laboratory of Analysis by Neutronic Activation of the RA-6 reactor is to determine the burn-up in fuels of the MTR type. In order to gain experience, up to the arrival of the hyperpure Germanium detector (HPGe) to be used in normal operation, preliminary measurements with a crystal of INa(Tl) were made. The fuel elements used are originated in the RA-3 reactor, with a decay superior to the thirteen years. For this reason, the unique visible photoelectric peak is the one of Cs-137, owing to the low resolution of the INa(Tl). After preliminary measurements, the profiles of burn-up, rectified by attenuation, were measured. Once the efficiency of the detector was determined, the calculation of the burn-up was made; for the element No. 144, a value of 21.6 ± 2.9 g was obtained to be compared with the value 21.9 g which was the evaluation made by the operators. (Author) [es

  20. Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

    2017-03-15

    In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  1. Experimental study associated to irradiation of FBR structural material, (4)

    International Nuclear Information System (INIS)

    1976-01-01

    The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

  2. Study on commercial FBR concepts by combining innovative technologies

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.; Kuroha, M.; Hida, T.

    1992-01-01

    A study was conducted on future prospects of FBR commercialization. Targets of further improving safety and economy were set to make commercial power plants that would be superior to future LWRs. Promising innovative technologies studied domestically and overseas were extracted by evaluating prospects for commercialization, effect, and plant applicability. Several commercial plants were conceptualized by introducing such technology to large-scale and oxide-fuel reactors. Estimates of construction cost, etc., proved that the targets could be achieved. A concept of long-term technological development was synthesized. (author)

  3. A study of the effects of changing burn-up and gap gaseous compound on the gap convection coefficient (in a hot fuel pin) in VVER-1000 reactor

    International Nuclear Information System (INIS)

    Rahgoshay, M.; Rahmani, Y.

    2007-01-01

    In this article we worked on the result and process of calculation of the gap heat transfer coefficient for a hot fuel pin in accordance with burn-up changes in the VVER-1000 reactor at the Bushehr nuclear power plant (Iran). With regard to the fact that in calculating the fuel gap heat transfer coefficient, various parameters are effective and the need for designing a model is being felt, therefore, in this article we used Ross and Stoute gap model to study impacts of different effective parameters such as thermal expansion and gaseous fission products on the h gap change rate. Over time and with changes in fuel burn-up some gaseous fission products such as xenon, argon and krypton gases are released to the gas mixture in the gap, which originally contained helium. In this study, the composition of gaseous elements in the gap volume during different times of reactor operation was found using ORIGEN code. Considering that the thermal conduction of these gases is lower than that of helium, and by using the Ross and Stoute gap model, we find first that the changes in gaseous compounds in the gap reduce the values of gap thermal conductivity coefficient, but considering thermal expansion (due to burn-up alterations) of fuel and clad resulting in the reduction of gap thickness we find that the gap heat transfer coefficient will augment in a broad range of burn-up changes. These changes result in a higher rate of gap thickness reduction than the low rate of decrease of heat conduction coefficient of the gas in the gap during burn-up. Once these changes have been defined, we can proceed with the analysis of the results of calculations based on the Ross and Stoute model and compare the results obtained with the experimental results for a hot fuel pin as presented in the final safety analysis report of the VVER-1000 reactor at Bushehr. It is noteworthy that the results of accomplished calculations based on the Ross and Stoute model correspond well with the existing

  4. Dynamic buckling and nonlinear response of FBR main vessels under earthquake loading

    International Nuclear Information System (INIS)

    Hagiwara, Yutaka; Kawamoto, Yoji; Nakagawa, Masaki; Akiyama, Hiroshi.

    1991-01-01

    Pseudo-dynamic tests of cylindrical shells under high temperature were performed in order to study elasto-plastic shear-bending buckling and the nonlinear response of FBR main vessels under earthquake loading. The test results showed a response reduction effect due to pre-buckling plasticity, and a large seismic margin due to post-buckling energy absorption of the cylinders. A simple expression of the response reduction effect was proposed, as a contribution to the safe and effective seismic design of FBRs. Two methods for seismic margin evaluation were also proposed, and it was shown that appropriate seismic margins can be ensured, when the response reduction effect is incorporated into the seismic design. (author)

  5. Feasibility study of application of ductless fuel assembly to FBR

    International Nuclear Information System (INIS)

    Itoh, K.; Shibahara, I.

    1996-01-01

    Feasibility studies on an application of the ductless fuel concept to an FBR core have been carried out in order to evaluate the basic features of the ductless core, especially in the fields of the thermal-hydraulic aspects and the mechanical behaviors. Regarding thermal-hydraulic aspects, analyses were performed by using a whole core thermal-hydraulic analysis code by making some modification for this study on the PWR code THINC. A small scaled ductless core model was prepared and a hydraulic experiment was carried out to study basic hydraulic characteristics of a ductless core. Core mechanical behaviors were analyzed focusing on the core irradiation bowing aspects and the seismic behaviors. Following features are revealed on the core structural behaviors: (1) the bowing stiffness of the ductless assembly is around 1/5 to 1/10 of that of the duct type assembly; (2) the contact loads between assemblies by the bowing effects are small through core cycles; (3) the damping of the ductless assemblies are so large that the seismic responses are small and the loads between assemblies are small due to occurring many contact points. Through this study it is expected that the concept of the ductless fuel can be applicable to FBR cores from the design view points of thermal-hydraulic and core mechanical behaviors

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

    International Nuclear Information System (INIS)

    Inagaki, Tatsutoshi

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Nakagiri, Toshio; Aoto, Kazumi; Hoshiya, Taiji

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Suzuki, T.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Braun, C.; Zebroski, E.L.

    1985-01-01

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

  11. Fuel assemblies for FBR type reactor

    International Nuclear Information System (INIS)

    Ikeda, Kiyoshi.

    1981-01-01

    Purpose: To decrease errors in the flow rate distribution of coolants by resiliently inserting a flow regulation rod having a variable flow regulation element formed at the upper portion along the axial direction in the entrance nozzle of a fuel assembly. Constitution: A plurality of orifice aperture are formed to the entrance nozzle of a fuel assembly and an aperture for inserting a flow regulation rod is formed to the top end of the entrance nozzle. A fixed flow regulation element A and a variable flow regulation element B supported coaxially with the nozzle by a support ring are disposed to the inside of the nozzle. The element B is urged by the resilient urging spring to the element A and connected by way of support lever to the flow regulation rod. While on the other hand, the top end of the nozzle is inserted through the partition wall between a high pressure coolant chamber and a low pressure coolant chamber. An aperture for hydrodynamically supporting the fuel assembly is provided by way of a frame and a flow regulation rod that stands vertically from the low pressure coolant chamber is disposed to the center of the frame. In the fuel assembly, the flow regulation rod inserted from the aperture at the top end of the nozzle pushes the element B upwardly to thereby maintain a flow passage of the coolant between the elements A and B. (Seki, T.)

  12. Main lessons for FBR safety study from the CABRI experiments

    International Nuclear Information System (INIS)

    Sato, Ikken

    2006-01-01

    CABRI project has been carried out FBR safety study with international cooperation of five nations since 1978. The project consists of four periods such as CABRI-1, CABRI-2, CABRI-FAST and CABRI-RAFT. The objects and the main information for hypothetical core decay accident and safety study of fuel are described. The behavior of core decay accident was studied in the first period (CABRI-1), the safety study of fuel was investigated after the second period (CABRI-2). Change of phenomena at the initial process of core decay accident, comparison between analysis and data of fuel diffusion behavior of neutron horoscope by CABRI-2 experiment, representative in-core experiment under slow TOP conditions, the damaged and undamaged pin under slow TOP conditions, the exterior of CABRI and TREAT and the upper part of CABRI and TREAT reactor are shown. CABRI experiments changed to LWR safety study and a part of TREAT is stated. (S.Y.)

  13. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  14. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Ohashi, Yukio.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Miyaji, Nobuyoshi.

    1979-01-01

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

  16. BWR noise spectra and application of noise analysis to FBR

    International Nuclear Information System (INIS)

    Nomura, T.

    1975-01-01

    Work related to noise analysis, in Tokyo Shibaura Electric Co. Ltd. (Toshiba) and Nippon Atomic Industry Group Co. Ltd. (NAIG) for the past several years is reviewed. After considering the Japan-United States Seminar on Reactor Noise Analysis in 1968, other subjects discussed were boiling water reactor noise analysis and work in relation to FBR. Parts of these are related to each other. For example, boiling detection and temperature fluctuations are problems pertinent to both fields. As the main problems in zero-power-reactor noise are now basically understood, only a brief description of the experiments involving the advanced two detector method is made. Focus is rather placed on the area of power plant noise. (author)

  17. Development of dose rate estimation system for FBR maintenance

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  18. Development of dose rate estimation system for FBR maintenance

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. Chinese FBR program and its first FR conceptual design

    International Nuclear Information System (INIS)

    Xu Mi

    1991-01-01

    As the preliminary study results, The long term strategy of FBR development in China would be divided into three steps: Experimental Fast Reactor, Modulized Fast Breeder Reactor and Large Fast Breeder Reactor. The design requirements, main technical selections and design boundary conditions for the first step, experimental fast reactor, which is named FFR (First Fast Reactor) in China have been given in the paper. The conceptual design of FFR, based on core neutronics, core subassemblies thermo-hydraulics, reactor block and mechanics, fuel handling system, main heat transfer system and steam-electricity generation system has been completed and also presented in this paper. Finally, the R and D program of FFR which is carring on and will be done is briefly introduced. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Nghiem, NP

    2003-11-16

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  2. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    2003-10-01

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

  6. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    Energy Technology Data Exchange (ETDEWEB)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V. [Inst. of Radiobiology, Minsk Univ. (Belarus); Boulyga, S.F. [Inst. of Inorganic Chemistry and Analytical Chemistry, Johannes Gutenberg-Univ. Mainz, Mainz (Germany); Becker, J.S. [Central Div. of Analytical Chemistry, Research Centre Juelich, Juelich (Germany)

    2005-07-01

    An analytical method is described for the estimation of uranium concentrations, of {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10{sup -9}g/g to 2.0 x 10{sup -6}g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and the average value amounted to 9.4{+-}0.3 MWd/(kg U). (orig.)

  7. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Brémier, S., E-mail: stephan.bremier@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Inagaki, K. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Capriotti, L.; Poeml, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Ogata, T.; Ohta, H. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany)

    2016-11-15

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15–20 μm and migration of cladding elements to the fuel. - Highlights: • Electron Probe MicroAnalysis of a UPuZr metallic fuel alloy irradiated to 7.0 at.% burn-up. • Significant redistribution of the fuel components along the fuel radius, nearly complete depletion of Zr in the central part of the fuel. • Interactions between the fuel and the cladding with occurrence of a number of intermediary layers and migration of cladding elements to the fuel. • Safe irradiation behaviour of the base alloy fuel.

  8. Optimization of FBR fuel element for high burnup

    International Nuclear Information System (INIS)

    Marbach, G.; Millet, P.

    1985-03-01

    After a brief historical background showing evolution of the French fast reactor fuel element from RAPSODIE to PHENIX and SUPER PHENIX we have examined the main points which have permitted to increase irradiation performance of the subassembly

  9. The MOX fuel behaviour test IFA-597.4/.5. Temperature and pressure data to a burn-up of 15 MWd/kg MOX

    International Nuclear Information System (INIS)

    Takano, K.

    1999-04-01

    The behaviour of MOX fuel should be investigated in detail for more effective use in the future, especially concerning its thermal performance and fission gas release. IFA-597.4 and IFA-597.5, containing two MOX fuel rods each with a fuel centre thermocouple and a pressure transducer, have been irradiated in the Halden Reactor to study the temperature threshold of fission gas release for MOX fuel and to explore potential differences in the thermal and fission gas release behaviour between solid and hollow pellets. The two rods of MOX fuel with an initial Pu-fissile content of 6.07 percent have solid pellets and hollow pellets respectively, and with an active length of about 220 mm. The diameter of the pellets is 8.05 mm with 180μm of diametral gap to the cladding. For the purpose of the test, power ramp operation, in which estimated peak temperature of the MOX pellets increases and decreases above and below the threshold for fission gas release in UO 2 fuel, is planned every 10 MWd/kgMOX of burn-up. The first ramp operation has been successfully performed at 10 MWd/kgMOX. When the estimated peak temperature of the fuel gets close to but below the threshold of UO 2 , fission gas release was observed at around 28 kW/m of power. Densification of the MOX pellets could be estimated to about 1.2 percent for the solid pellets and about 2,3 percent for the hollow pellets from normalised internal rod pressure. After 13.5 MWd/kgMOX the average assembly power has been operated low enough to observe swelling rate of MOX fuel pellets and behaviour after significant fission gas release. The burn-up had reached 15.5 MWd/kgMOX as of the end of 1998. The target burn-up of this MOX test is 60 MWd/kgMOX (author) (ml)

  10. Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel and NDE assessment

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Kim, Jong Bum; Kim, Seok Hun; Yoo, Bong

    2001-03-01

    Creep fatigue crack growth contributes to the failure of FRB reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety and setup the in-service inspection strategy. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS crack, NDE short crack and shallow long crack) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. The total crack growth of the largest PTS crack is very small after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life. The crack depth growth of the shallow long crack is faster than that of the NDE short crack. In the ISI of the large-scale FBR reactor vessel, the ultrasonic inspection is beneficial to detect the shallow circumferential cracks.

  11. Burn-up credit criticality safety benchmark phase VII - UO2 fuel: study of spent fuel compositions for long-term disposal

    International Nuclear Information System (INIS)

    2012-01-01

    After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments. This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (k eff ) values for pressurised-water-reactor (PWR) UO 2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and k eff values have been calculated for 30 post-irradiation time steps out to one million years

  12. Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-01-01

    Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well

  13. The MOX Fuel Behaviour Test IFA-597.4: Temperature And Pressure Data To A Burn-Up Of 5.4 MWd/kg MOX

    International Nuclear Information System (INIS)

    McGrath, M. A.; Teshima, H.

    1998-02-01

    Characterising the behaviour of MOX fuel is becoming increasingly important as many commercial reactors are or will be operating with this type of fuel. With this as a driving force, a new joint programme experiment, IFA-597.4, has been loaded into the reactor at Halden for the purpose of establishing the fission gas release behaviour of MOX fuel. Both annular and solid pellet fuel is being utilised and the irradiation is being conducted such that the fuel is initially operated below the onset of fission gas release. The fuel will later be subjected to small power up ratings which will be held for short periods of time. These are designed to bring the fuel to just above the temperature threshold for fission gas release thus allowing the FGR behaviour of both solid and annular MOX fuel to be established. The rig contains two fuel rods of active length 220 mm and diameter 8.05 mm. Both fuel rods contain MOX fuel with an initial Pu-fissile content of 6.07% and both are instrumented with a fuel centre thermocouple and a pressure transducer. The test is being performed under HBWR conditions and at the time of the reactor shutdown at the end of 1997 a mean burn-up of 5.4 MWd/kg MOX had been achieved with the rods at an average rating of 30 kW/m. The rod pressure data show that no fission gas had been released up to the shutdown. The fuel centre temperatures of both rods exhibit an initial increase concurrent with a fall in the monitored rod internal pressures as a result of fuel densification. It was estimated that about 1-1.4% fuel densification by volume had occurred in the two rods by a burn-up of about 3 MWd/kg MOX. (author)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  15. Temperature fluctuation reducing device for FBR type reactor

    International Nuclear Information System (INIS)

    Ootsuka, Fumio; Shiratori, Fumihiro.

    1991-01-01

    In existent FBR type reactors, since temperature fluctuation in the reactor upper portion has been inevitable, thermal fatigue may be caused possibly in reactor core upper mechanisms. Then, a valve is disposed to a control rod lower guide tube contained in a reactor container for automatically controlling the amount of passing coolants in accordance with the temperature of the passing coolants, to mix and control coolants passing through a fuel assembly in adjacent with the guide tube and coolants passing through the guide tube. Further, a rectification cylinder is disposed, in which a portion of coolants passing through the fuel assembly is caused to flow. An orifice is disposed to the cylinder with an exit being disposed to the upstream thereof such that the coolants not flown into the rectification cylinder and the coolants passing through the guide tube are mixed to moderate the temperature fluctuation. That is, a portion of the coolants flown into the rectification cylinder can not pass through the orifice, but flow backwardly to the upstream and is discharged out of the rectification cylinder from the coolants exit and mixed sufficiently with coolants passing through the guide tube. In this way, temperature fluctuation can be moderated. (N.H.)

  16. Development of expert system for structural design of FBR components

    International Nuclear Information System (INIS)

    Ueda, Hiroyoshi; Uno, Masayoshi; Ogawa, Hiroshi; Shimakawa, Takashi; Yoshimura, Shinobu; Yagawa, Genki.

    1995-01-01

    The characteristics of structural design processes for nuclear components can be summarized as follows : (1) Many engineers belonging to different fields are working in parallel, exchanging a huge amount of data and information. (2) A final solution is determined after a number of iterative design processes. (3) Solutions have to be examined many times based on sophisticated design codes. (4) Sophisticated calculation methods such as the finite element method are frequently utilized, and experts' knowledge on such analyses plays important roles in the design process. Taking these issues into consideration, a new expert system for structural design is developed in the present study. Here, the object-oriented data flow mechanism and the blackboard model are utilized to systematize structural design processes in a computer. An automated finite element calculation module is implemented, and experts' knowledge is stored in knowledge base. In addition, a new algorithm is employed to automatically draw the design window, which is defined as an area of permissible solutions in a design parameter space. The developed system is successfully applied to obtain the design windows of four components selected from the demonstration FBR structures. (author)

  17. Development of FBR integrity system code. Basic concept

    International Nuclear Information System (INIS)

    Asayama, Tai

    2001-05-01

    For fast breeder reactors to be commercialized, they must be more reliable, safer, and at the same, economically competitive with future light water reactors. Innovation of elevated temperature structural design standard is necessary to achieve this goal. The most powerful way is to enlarge the scope of structural integrity code to cover items other than design evaluation that has been addressed in existing codes. Items that must be newly covered are prerequisites of design, fabrication, examination, operation and maintenance, etc. This allows designers to choose the most economical combination of design variations to achieve specific reliability that is needed for a particular component. Designing components by this concept, a cost-minimum design of a whole plant can be realized. By determining the reliability that must be achieved for a component by risk technologies, further economical improvement can be expected by avoiding excessive quality. Recognizing the necessity for the codes based on the new concept, the development of 'FBR integrity system code' began in 2000. Research and development will last 10 years. For this development, the basic logistics and system as well as technologies that materialize the concept are necessary. Original logistics and system must be developed, because no existing researches are available in and out of Japan. This reports presents the results of the work done in the first year regarding the basic idea, methodology, and structure of the code. (author)

  18. Hydrogen transfer preventive device in FBR power plant

    International Nuclear Information System (INIS)

    Hoshi, Yuichi.

    1987-01-01

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

  19. Mass transfer of steels for FBR in sodium loop

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  20. Improvement of numerical analysis method for FBR core characteristics. 3

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yamamoto, Toshihisa; Kitada, Takanori; Katagi, Yousuke

    1998-03-01

    As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method; A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. Part 2: Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory; Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. Part 3: Improvement of Nodal Transport Calculation for Hexagonal Geometry; A method to evaluate the intra-subassembly power distribution from the nodal averaged neutron flux and surface fluxes at the node boundaries, was developed based on the transport theory. (J.P.N.)

  1. Material attractiveness of plutonium composition on doping minor actinide of large FBR

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Kuno, Yusuke

    2011-01-01

    Material attractiveness analysis on isotopic plutonium compositions of fast breeder reactors (FBR) has been investigated based on figure of merit (FOM) formulas as key parameters as well as decay heat (DH) and spontaneous fission neutron (SFN) compositions. Increasing minor actinide (MA) doping gives the significant effect to increase Pu-238 composition. However, the compositions of Pu-240 and Pu-242 become less with increasing MA doping. DH and SFN compositions in the core regions similar to the DH and SFN compositions of MOX-grade. Material attractiveness based on FOM1 formula shows all isotopic plutonium compositions in the blanket regions as well as in the core regions are categorized as high attractive material. Adopted FOM2 formula can distinguishes the material attractiveness levels which show the plutonium compositions in blanket regions as high attractiveness level and its composition in the core regions as low level of material attractiveness. MA doping is effective to reduce the material attractiveness level of blanket regions from high to medium and it requires much more MA doping rate to achieve low level of attractiveness (FOM<1) based on adopted FOM1 formula. Low material attractiveness level can be obtained by 4 % or more doping MA based on adopted FOM2 formula which considers not only DH composition effect, but also SFN composition effect that gives relatively higher contribution to material barrier of plutonium isotopes. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  3. Recent R/D towards aqueous reprocessing of FBR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Mallika, C.; Pandey, N.K.; Kumar, S.; Kamachi Mudali, U. [Materials, Process and Equipment Development Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-07-01

    The mixed Pu-rich carbide spent fuel with a burn up of 155 GWd/t from the Fast Breeder Test Reactor is being reprocessed in a hot-cell facility by PUREX process. Based on the input from the operation of this facility, research and development activities were carried out to improve the recovery, decontamination factors, economy and to reduce the waste volumes. Reduction of uranyl ions in a continuous flow electrochemical reactor and electrolytic as well as chemical reduction of 4 M HNO{sub 3} from liquid waste could be performed in continuous mode. Using the optimized parameters, suitable electrolytic cells/experimental setups were designed for the plant capacity of 6 L/h. Studies on the extraction kinetics of Ru with 30% TBP (tributyl phosphate) in NPH revealed that better decontamination factor with respect to Ru can be achieved using fast contactors like centrifugal extractors (CEs). Towards developing a spent solvent recovery system to reduce organic waste volumes, a pilot plant was set up, which could recover diluent as top product of distillation column and 40% TBP as bottom product from inactive degraded solvent. A solvent recovery system using short path distillation was also developed for installation in hot cells. (authors)

  4. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    International Nuclear Information System (INIS)

    Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

    2014-01-01

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice

  5. Application of instrumental neutron activation analysis of uranium in burn-up measurements using. gamma. -ray spectrometric method

    Energy Technology Data Exchange (ETDEWEB)

    Chao, H E; Lu, W D

    1975-12-01

    In uranium burnup measurements, the amount of uranium in the irradiated sample needs to be determined, and the application of instrumental neutron activation analysis for this purpose is investigated. The method uses the gamma-ray activities of /sup 239/Np and some short-lived fission products of half-lives no longer than a few days to determine the quantities of /sup 238/U and /sup 235/U respectively. The advantages of the method include: (1) the amounts of both /sup 235/U and /sup 238/U of the sample can be simultaneously determined with good accuracy, (2) the same sample may be used to determine both the fission numbers and the amount of uranium remaining simultaneously or one after another, thus the exact amount of the sample is not necessarily known, (3) since the amount of the sample needed for the determination is usually small, i.e., about 10 ..mu..g, it should be easily handled even for high-level burnup samples. The error of the method is about 3 percent for a single measurement. The burnup values measured for an irradiated natural uranium sample from three aliquots using several fission products are in good agreement. The effective cross section for /sup 235/U deduced from the burnup and the integrated flux from a cobalt monitor is found to be 589 +- 19 barn which is in agreement with the literature value of 577 +- 1 barn.

  6. Measurement of gamma attenuation coefficients in UO2 and zirconium for self-absorption corrections of burn-up determination

    International Nuclear Information System (INIS)

    Podest, M.; Klima, J.; Stecher, P.; Stecherova, E.

    1978-01-01

    UO 2 pellets from ALUOX fuel elements were used in measuring the absorption coefficient of gamma radiation in UO 2 . The results of measurements of the energy dependence of the linear absorption coefficient (within 622 to 796 keV) and of the dependence on pellet density showed that in the given density interval the absorption coefficient was almost constant. The density interval was chosen to be typical for pellet fuel used in water cooled and water moderated power reactors. The results are also shown of the dependence of the mass absorption coefficient of gamma radiation in Zr on radiation energy and compared with the mass absorption coefficient of Mo; these also showed the independence of the absorption coefficient on density. The linear and mass absorption coefficients of UO 2 are considerably high and correspond approximately to the absorption coefficient of lead. For the measured energy range the variation of absorption coefficient is about 40%, which causes errors in burnup determination. The efficiency was also determined of Ge(Li) detectors for the energy range 0.5 to 1.2 MeV. The determination of the above coefficients was used for improving the gamma fuel scanning technique in determining the activity and burnup of spent fuel elements. (J.P.)

  7. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    International Nuclear Information System (INIS)

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-01-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  8. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  11. Development of end plug welding method in the fabrication of FBR fuel pins

    International Nuclear Information System (INIS)

    Ohtani, Seiji; Sawayama, Takeo; Tateishi, Yoshinori

    1977-01-01

    As a part of the development of the automatic and remote controlled fabrication of FBR fuel pins, welding of fuel pin end plugs has been examined. Cladding tubes and end plugs used for this experiment are made of SUS 316, and they are the components of fuel pins for the prototype fast breeder reactor (Monju) or the second core of Joyo (Joyo MK-II). The welding tests of cladding tubes and four kinds of end plugs were carried out by means of two techniques; tungsten inert gas welding and laser welding. It can be said that no considerable difference was observed in weld penetration, occurrence rate of weld defects and breaking strength between the tight fit and the loose fit plugs. The face-to-face fit welding requires the least welding heat input, but involves much difficulty in the control of weld penetration and bead zone diameter. The good concentrative property and high energy density of laser beam make the face of weld hollow due to the vaporization of weld metal. However, this problem can be easily solved by changing the shape of end plugs. Good results in the other characteristics of the weld also were obtained by this laser welding. Further experiment is needed in connection with the compatibility of weld metal with sodium and neutron irradiation before final judgement is made on the laser welding technique. (Nakai, Y.)

  12. Visualization test using piping group mock up specimen for evaluation of wastage phenomena in steam generator for FBR

    International Nuclear Information System (INIS)

    Kato, Keisuke; Yoshida, Atsuro; Arae, Kunihiko; Narabayashi, Tadashi; Ohshima, Hiroyuki; Kurihara, Akikazu

    2012-01-01

    There is a need for quantitative evaluation of wastage phenomena in steam generator for FBR. We focused attention on liquid droplet impingement erosion (LDIE) in wastage phenomena and performed basic study with piping group mock up specimen for quantitative evaluation of LDIE. First, we did visualization test of high pressure and high speed jet into the water. Test section mock up the crack of heat exchanger tube and neighboring heat exchanger tubes. We did the test under the following test conditions. Upstream pressure is 0.3MPa, vapor temperature is 300K, crack width is 0.1mm, and crack length is 40mm. (crack diameter is 0.2mm) Second, we did pressure and temperature measurement test in the same test conditions as before. We evaluated jet behavior at test section by those two tests. In addition, we did two phase flow analysis of the jet with TRAC code. (author)

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

    International Nuclear Information System (INIS)

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

    1981-12-01

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

  14. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  15. A study on adsorption onto TODGA resin after electrolytic reduction in ERIX process for reprocessing spent FBR-MOX fuel

    International Nuclear Information System (INIS)

    Hoshi, Harutaka; Arai, Tsuyoshi; Wei, Yuezhou; Kumagai, Mikio; Asakura, Toshihide; Morita, Yasuji

    2005-01-01

    For reprocessing spent FBR-MOX fuel, an advanced aqueous reprocessing process ''ERIX process'' has been developed. In this system, hydrazine is used as reduction holding reagent for the valance adjustment of U by electrolytic reduction in nitric acid solution. Therefore, hydrazine is contained in high level liquid waste after separation of U, Pu and Np. Effect of hydrazine on adsorption of FP onto TODGA resin was examined. When hydrazine concentration was less than 0.3 M, effect on the distribution coefficient was negligibly small. After electrolytic reduction, some elements exist as lower valence state. Ru and Tc are most difficult elements to control their behavior in aqueous process. The distribution coefficient of both Ru and Tc onto TODGA decreased after electrolytic reduction, because they are reduced to lower valence. Hence, it is difficult for Ru or Tc to diffuse to allover the process and separation of MA from Tc and Ru was enhanced by electrolytic reduction. (author)

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

    International Nuclear Information System (INIS)

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

    2001-03-01

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

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

    International Nuclear Information System (INIS)

    Maekawa, Isamu; Kurata, Chikatoshi

    2004-02-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-06

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

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

    International Nuclear Information System (INIS)

    Kani, Yoshio

    2000-01-01

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

  1. JAEA FBR Plant Engineering Center annual report 2011

    International Nuclear Information System (INIS)

    2012-11-01

    The FBR Plant Engineering Center was established on April 1, 2009 located in a research building, of which care is taken by the International Nuclear Information Training Center, Tsuruga Head Office, at Shiraki in Tsuruga. The mission of the center is to perform R and D (research and development) works both for analysis of operational experiences at the prototype fast breeder reactor “Monju” and for technology development concerning design and operation of “Monju”. Moreover it is also required to apply the results to next generation fast breeder reactors, which is an important role of Advanced Nuclear System Research and Development Directorate. And in these R and D activities, it is expected to conduct the works in cooperation with domestic or foreign research organizations or universities by a joint-study or a collaborative-work manner. The R and D activities have been carried out specifically on the “demonstration of the reliability as a power generation plant” and “establishment of sodium handling technology”, which are originally intended missions of “Monju”. And the other R and Ds have been promoted both for the plant engineering, such as plant maintenance, to effectively use an existing reactor in order to apply the R and D results to a future demonstration reactor, and for the irradiation test study, such as advanced fuel irradiation, to use “Monju” as an irradiation test bed. In order to perform these R and D activities, five R and D groups have been set up in the center. They are operation-and-maintenance engineering, sodium engineering, reactor-core-and-fuel engineering, plant engineering, and safety engineering groups. However, the Japanese atomic energy policy is being reviewed after the accident of the Fukushima Daiichi nuclear power station caused by a tsunami generated by the Tohoku-district-off-the-Pacific-Ocean Earthquake on March 11, 2011, and all the R and D activities using “Monju” have been suspended since late 2011

  2. Some difference of concepts between design guideline for FBR base isolation system and aseismic design guideline of LWR in Japan

    International Nuclear Information System (INIS)

    Shibata, Heki

    1992-01-01

    This paper deals with the concept and the relation of 'the Base Isolation System and FBR' to the Safety Criteria and the Guideline of the Aseismic Design of LWR in Japan. The Central Research Institute of Electric Power Industries have been working for FBR last several years. The author has been contribute to their works, and this is one of the subjects. He described his own idea obtained through the cooperative work with CRIEPI. (author)

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

    International Nuclear Information System (INIS)

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

    1985-03-01

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

  4. A study on the design method of the laminated rubber bearing for FBR

    International Nuclear Information System (INIS)

    Mazda, T.; Ishida, K.; Ikeuchi, K.; Yashiro, T.

    1994-01-01

    The Central Research Institute of Electric Power Industry (Japan) has been carrying out the Demonstration Test and Research Program of the Seismic Isolation System for the Fast Breeder Reactor (FBR) under contract with the Ministry of International Trade and Industry (FY1987--FY1993). In this research program, development of the seismic isolation element, reliability evaluation of the seismic isolation system, and examination of a design-based earthquake have been conducted in order to develop and propose design guidelines of seismic isolation system for FBR plant. The purpose of this paper is to describe the test results concerned with structural design of the laminated rubber bearing, and to explain the construction of the design method for the laminated rubber bearing, which is one of the most important factors in the design guidelines of the seismic isolation system

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

    International Nuclear Information System (INIS)

    Nakagawa, Y.

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

  6. Thermal Neutron Capture and Thermal Neutron Burn-up of K isomeric state of 177mLu: a way to the Neutron Super-Elastic Scattering cross section

    International Nuclear Information System (INIS)

    Roig, O.; Belier, G.; Meot, V.; Daugas, J.-M.; Romain, P.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Letourneau, A.; Marie, F.; Veyssiere, Ch.

    2006-01-01

    Thermal neutron radiative capture and burn-up measurements of the K isomeric state in 177Lu form part of an original method to indirectly obtain the neutron super-elastic scattering cross section at thermal energy. Neutron super-elastic scattering, also called neutron inelastic acceleration, occurs during the neutron collisions with an excited nuclear level. In this reaction, the nucleus could partly transfer its excitation energy to the scattered neutron

  7. Burn-Up Determination by High Resolution Gamma Spectrometry: Spectra from Slightly-Irradiated Uranium and Plutonium between 400-830 keV

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Ronqvist, N

    1966-08-15

    Previously published studies of the short-cooled fission product spectra of irradiated uranium have been severely restricted by the poor energy resolution of the sodium iodide detectors used. In this report are presented fission product spectra of irradiated uranium and plutonium obtained by means of a lithium-drifted germanium detector. The resolved gamma peaks have been assigned to various fission products by correlation of measured energy and half-life values with published data. By simultaneous study of the spectra of two irradiated mixtures of plutonium and uranium, the possibility of using the activities of Ru-103 and Ru-106 as a measure of the relative fission rate in U-235 and Pu-239 has been briefly examined.

  8. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    International Nuclear Information System (INIS)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F.

    2000-01-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  9. Burn-Up Determination by High Resolution Gamma Spectrometry: Spectra from Slightly-Irradiated Uranium and Plutonium between 400-830 keV

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Ronqvist, N.

    1966-08-01

    Previously published studies of the short-cooled fission product spectra of irradiated uranium have been severely restricted by the poor energy resolution of the sodium iodide detectors used. In this report are presented fission product spectra of irradiated uranium and plutonium obtained by means of a lithium-drifted germanium detector. The resolved gamma peaks have been assigned to various fission products by correlation of measured energy and half-life values with published data. By simultaneous study of the spectra of two irradiated mixtures of plutonium and uranium, the possibility of using the activities of Ru-103 and Ru-106 as a measure of the relative fission rate in U-235 and Pu-239 has been briefly examined

  10. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F

    2000-07-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  11. Developing maintenance technologies for FBR's heat exchanger units by advanced laser processing

    International Nuclear Information System (INIS)

    Nishimura, Akihiko; Shimada, Yukihiro

    2011-01-01

    Laser processing technologies were developed for the purpose of maintenance of FBR's heat exchanger units. Ultrashort laser processing fabricated fiber Bragg grating sensor for seismic monitoring. Fiber laser welding with a newly developed robot system repair cracks on inner wall of heat exchanger tubes. Safety operation of the heat exchanger units will be improved by the advanced laser processing technologies. These technologies are expected to be applied to the maintenance for the next generation FBRs. (author)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Yokoo, Takeshi; Inoue, Tadashi

    1999-01-01

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

  15. Report of the evaluation by the project evaluation committee on research and development of FBR structural materials. Result intermediate evaluation in fiscal year 2002

    International Nuclear Information System (INIS)

    2003-03-01

    The project evaluation committee of FBR (Fast Breeder Reactor) and fuel cycle evaluated the problem under the evaluation method defined by the committee on the basis of material and supplementary data produced by JNC (Japan Nuclear Cycle Development Institute) and discussion. This report consists of summary, construction of the committee, investigation process, evaluation method, opinion of object, plan, research system, results, relation to practical use, public offering of results, development and others and total evaluation. Each opinion of the committee member are reported. Development of structural material is important problem of R and D of FBR. The estimation techniques of materials, high temperature structural and earthquake-resistant design techniques are evaluated to obtain the fruitful results. The remarkable results are development of materials of 316FR, using plate to reactor, estimation method of thermal striping, improvement of earthquake-resistant analysis technology, contribution to evaluation analysis of Monju second liner and basic information to develop quantity technique of material damage. (S.Y.)

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

    International Nuclear Information System (INIS)

    Hara, Hideaki

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-07-01

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

  18. Engineering scale tests of mechanical disassembly and short stroke shearing systems for FBR fuel assembly

    International Nuclear Information System (INIS)

    Higuchi, Hidetoshi; Kitagaki, Toru; Koizumi, Kenji; Hirano, Hiroyasu; Takeuchi, Masayuki; Washiya, Tadahiro; Kawabe, Yukinari; Kobayashi, Tsuguyuki

    2011-01-01

    Japan Atomic Energy Agency (JAEA) and The Japan Atomic Power Company (JAPC) have been developing an advanced head-end process based on mechanical disassembly and short stroke shearing systems as a part of Fast Reactor Cycle Technology Development (FaCT). Fuel pins for a fast reactor are installed within a hexagonal shaped wrapper tube made of stainless steel. In order to reprocess the fast reactor fuel pins, they must be removed from the wrapper tube and transported to the shearing system without failure. In addition, the advanced aqueous reprocessing process, called 'NEXT' (New Extraction System for TRU Recovery) process requires a solution of the spent fuel with relatively high concentration (500g/L). JAEA and JAPC have developed the mechanical disassembly and the short stroke shearing technology which is expected to make fragmented fuel to satisfy these requirements. This paper reports the results of engineering scale tests on the mechanical disassembly and short stroke shearing systems. These tests were carried out with simulated FBR fuel assembly and removed pins. The mechanical cutting method has been developed to avoid fuel pin failure during disassembly operation. The cutting process is divided into two modes, so called 'slit-cut' for cutting the wrapper tube and 'crop-cut' for the end plug region of the fuel pin bundle. In the slit-cut mode, the depth of cutting was automatically controlled based on the calculated wastage of the cutting tool and deformation of the wrapper tube which had been measured before the cutting. This procedure was confirmed to minimize the fuel pin failure which was hard to prevent in the case of laser cutting. The cutting speed was also controlled automatically by the electric current of the cutting motor to lower the load of the cutting tool. The removed fuel pins were transported to the shearing machine, whose fuel shearing magazine width was set to be narrow to realize the suitable configuration for the short stroke shearing

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

    International Nuclear Information System (INIS)

    Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Ishikawa, Makoto

    1994-01-01

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

  2. Present condition of survey research on actualization strategy of fast breeder reactor (FBR) cycling. Design research on fuel production system

    International Nuclear Information System (INIS)

    Tanaka, Kenya

    2001-01-01

    The fuel production system design investigation was performed for construction of fuel production process concept and plant image searching for the targets such as economics, environmental loading reduction, and so on required for practical use of FBR fuel recycling at a premise of safety security. By expectation of economics as a fuel cycling system, enhancement of nuclear proliferation resistance, and so on, it becomes more important to investigate on a fuel cycling system suitable for raw materials with low decontamination and high radiation intensity. In addition, it is also necessary to carry out investigation on fuel production system concept accompanies with MA recycling system for reduction of environmental loading. Therefore, investigation objects on the system were laid their fundamental processes on denitrification conversion/pelletizing process and gelation/vibration filling process for raw material solution from advancing wet reprocessing and on vibration filling process for oxide granules obtained from dry reprocessing system and casting method for metal fuels. As a result, for the pollution removal fuel production system suitable for either of wet/dry reprocessing, a mass-production scale production plant image was elucidated at a premise of production yield, realizability of remote automation system, and so on. On candidate concepts of every fuel production system, no fatal defect was found on results of outline evaluation on features of system such as production facility scale and so on before present stage. (G.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  5. Development of 3-D FBR heterogeneous core calculation method based on characteristics method

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Maruyama, Manabu; Hamada, Yuzuru; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro

    2002-01-01

    A new 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed by combining the characteristics method and the nodal transport method. In the axial direction the nodal transport method is applied, and the characteristics method is applied to take into account the radial heterogeneity of fuel assemblies. The numerical calculations have been performed to verify 2-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction rates in a small region

  6. LBB assessment on ferrite piping structure of large-scale FBR

    OpenAIRE

    兪 淵植

    2002-01-01

    These days, this interest on LBB(Leak before Break) design becomes to be rising in the viewpoint of the cost reduction and structural inter-grity for the commercialization of FBR plants, LBB design enables pla-nts to be shut down safely before occuring unstable fracture by dete- cting the leak rates even if a crack initiates and penetrates a wall thickness. It is necessary to assess crack growth and penetration be- havior considering in-service conditions under operation temperature, leak re...

  7. FBR Italian position. The PEC reactor and its control rods system

    International Nuclear Information System (INIS)

    Artioli, C.

    1984-01-01

    The greatest effort in Italy concerning Fast Breeder Reactors is concentrated on the PEC project. This project represents in fact the sole Italian work on the FBR option. Up to now the state of progress of the PEC project is as follows: - Design work on reactor components and related systems: 80% - Civil engineering work: 45% - Components construction: 18%. It has recently been stated that construction of the plant must be finished by 1987. The progressive design of the PEC control system and the supporting research and development programmes are described in this paper

  8. Studies in the dissolver off-gas system for a spent FBR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Heinrich, E.; Huefner, R.; Weirich, F.

    1982-01-01

    Investigations of possible modifications of the process steps of a dissolver off-gas (DOG) system for a spent FBR fuel reprocessing plant are reported. The following operations are discussed: iodine removal from the fuel solution; behaviour of NOsub(x) and iodine in nitric acid off-gas scrubbers at different temperatures and nitric acid concentrations; iodine desorption from the scrub acid; selective absorption of noble gases in refrigerant-12; cold traps. The combination of suitable procedures to produce a total DOG system is described. (U.K.)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  10. Documentation for WIMSD-formatted libraries based on ENDF/B-VII.1 evaluated nuclear data files with extended actinide burn-up chains and cross section data up to 2000 K for fuel materials

    International Nuclear Information System (INIS)

    López Aldama, Daniel

    2014-11-01

    In the frame of WIMS Library Update Project the WIMSD-IAEA-69 and WIMSD-IAEA-172 libraries were prepared and made available at the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The main libraries were prepared from different sources of evaluated nuclear data that were available before December 2003. Also others WIMSD libraries were prepared from the major evaluated nuclear data libraries and made available at http://www-nds.iaea.org/wimsd. During the last ten years new libraries have been prepared every time that a major version of an evaluated nuclear data library has been released, namely JEFF-3.1 and ENDF/B-VII.0. Recently, end-users have requested to extend the temperature ranges of fuel materials included in the libraries and also to extend the burn-up chains to higher actinides up to Cf-254. The inclusion of new structural materials, like bismuth, has been also considered. Therefore, new WIMSD-formatted libraries in the 69- and 172-energy structure have been prepared with more materials, extended actinides burn-up chains and higher temperatures in thermal and resonance range

  11. Development of FBR visual inspection technique in sodium

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-07-01

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

  13. Study on reprocessing plant during transition period from LWR to FBR

    International Nuclear Information System (INIS)

    Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

    2011-01-01

    We have proposed a concept of a reprocessing plant suitable for the transition period from the light water reactors (LWRs) to the fast breeder reactors (FBRs) by making comparison of two plant concepts: (1) Independent Plant which processes LWR fuel and FBR fuel in separately constructed lines and (2) Modularized Plant which processes LWR fuel and FBR fuel in a same line. We made construction plans based on the reference power generation plan, and evaluated the Pu supply capability using the power generation plan as an indicator of plant operation flexibility. In general, a margin of processing capacity increases the Pu supply capability. The margin of the Modularized Plant necessary to obtain equivalent Pu supply capability is smaller than that of the Independent Plant. Also the margin of the Independent Plant results in decrease in the plant utilization factor. But the margin of the Modularized Plant results in little decrease in the plant utilization factor, because the Modularized Plant can address the types of reprocessing fuel to adjust to Pu demand and processing capacity. Therefore, the Modularized Plant has a greater potential for the reprocessing plants during transition period. (author)

  14. Investigation of 144Pr measurements for determination of residual Pu in FBR leached hulls

    International Nuclear Information System (INIS)

    Yates, A.; Bremner, W.B.

    1983-12-01

    The measurement of plutonium in leached hulls arising from FBR fuel reprocessing is required for plant control, accountancy and safeguards. At DNPDE these measurements are carried out on the batch of hulls prior to bulking into 200 l drums for retrievable storage and ultimately further treatment for plutonium recovery. The experience to date has related to the use of neutron interrogation using sealed tube neutron generators as the irradiation source. The supply of replacement sealed tubes has become difficult. It was therefore decided to consider, amongst other techniques, the measurement of 144 Pr gamma emission as a possible alternative. The technique has been extensively used for thermal reactor fuels on a routine basis. There has also been a limited amount of work reported from Cap La Hague on applying the technique to FBR fuels as part of an experimental programme. This paper therefore describes the work done at DNPDE which evaluated the technique for use on a batch of hulls arising from a PFR fuel reprocessing campaign. (author)

  15. Roof slab cooling device in a FBR type reactor

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1987-01-01

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

  16. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

    A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

  17. Observation method for inside of FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio; Ishitori, Takashi.

    1992-01-01

    The method of the present invention provides such a method that the surface of a metal case of an ultrasonic transducer keep a good intimate contact with liquid sodium always in a normal state in a short period of time while exhibiting satisfactory wettability. That is, an oxygen concentration in liquid sodium is increased and the inside of the reactor is seen through. Liquid sodium in a state of high oxygen concentration has extremely satisfactory wettability with metals. Accordingly, the metal surface of the ultrasonic transducer can be put to a intimate contact with the liquid metal sodium in a normal state. Further, a coating layer made of nickel or gold is disposed on the surface of the ultrasonic transducer. With such a constitution, the wettability with the liquid metal sodium can further be improved. (I.S.)

  18. Gas supply during fluidization of spherical particles in FBR

    International Nuclear Information System (INIS)

    Jeong, Kyung Chai; Eom, Sung Ho; Kim, Yeon Ku; Kim, Woong Ki; Kim, Young Min; Lee, Young Woo; Cho, Moon Seong

    2011-11-01

    Calculations of gas flow requirements and of other related parameters in the fluidized-bed process used to coat nuclear fuel particles are presented. These data include: volumes and surfaces of spheres for diameters of 50 to 500μm: number of theses spheres in 1 g for densities of 2 to 11 g/cm 3 : overall densities of coated spheres for initial particle diameters of 50 to 500μm, initial densities of 8 to 11 g/cm 3 , coating densities of 1.0 to 2.2 g/cm 3 , and final particle diameters of 100 to 1000μm: viscosities of Ar, CO 2 , He, and H 2 for temperatures up to 2200 .deg. C: minimum flows of He and Ar necessary to fluidized nuclear fuel particles at 20 .deg. C: coefficients for converting the 20 .deg. C minimum fluidization gas flows to high-temperature flows (up to 2200 .deg. C): variation of particle diameter with time for constant weight deposition rate: variation of coating gas flow for constant linear growth of the coating: comparison of coating time at constant weight deposition rate and at constant coating growth rate

  19. Irradiation behavior of graphite shielding materials for FBR

    International Nuclear Information System (INIS)

    Maruyama, Tadashi; Kaito, Takeji; Onose, Shoji; Shibahara, Itaru

    1994-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor 'JOYO' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  2. Development of centrifugal contactor for FBR fuel reprocessing

    International Nuclear Information System (INIS)

    Washiya, Tadahiro; Takeuchi, Masayuki; Suganuma, Takashi; Aose, Shinichi; Ogino, Hideki

    2003-01-01

    In the Feasibility Study on Commercialized Fast Reactor Cycle Systems, the aqueous reprocessing technology is nominated as a candidate for future reprocessing system, which supposes to apply a centrifugal contactor in the extraction process. For the reprocessing plant, the centrifugal contactor has great advantages such as reducing solvent degradation, improving of equipment utilization rate, compact designing of equipment layout and critical safety domination. From these advantages, the centrifugal contactor is crucial equipment in the aqueous reprocessing process. Since 1985, JNC has been developing the centrifugal contactor. The single unit development has been accomplished and basic characteristics such as extraction performance, fluidic performance and remote maintenance performance have been determined. A durability test has been conducted for high longevity, with consideration given to the nitric acid mist and estimation of the equipment lifetime. System test equipment with centrifugal contactors of engineering scale was installed, and uranium test was conducted. Up to now, a standard flow sheet test in the extraction process and mal-operation test assuming the one stage shutdown condition have been performed. (author)

  3. Sodium leakage experience at the prototype FBR Monju

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  4. Method of determining the composition of fuels for FBR type reactors

    International Nuclear Information System (INIS)

    Tsutsumi, Kiyoshi.

    1981-01-01

    Purpose: To improve the core safety of FBR type reactors by determining the composition of fuels composed of oxide mixture of plutonium and uranium, using a relation between specific plutonium seed and plutonium enrichment degree. Method: Relation is determined between the ratio of a specific plutonium seed for constituting plutonium oxide, for example 239 U ratio and a plutonium enrichment degree required for setting the assembly power to a constant level. The ratio of 239 U is plutonium having a given isotopic ratio is also determined. The accuracy of the 239 U ratio can be improved by the correction using the density coefficient. Then, the plutonium enrichment degree is determined using the relation determined as above based on the thus determined 239 U ratio. The composition of the fuel using oxide mixture of plutonium and uranium is determined by utilizing the thus obtained plutonium enrichment degree. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  6. Simulation and experimental approach to CVD-FBR aluminide coatings on ferritic steels under steam oxidation

    International Nuclear Information System (INIS)

    Leal, J.; Alcala, G.; Bolivar, F.J.; Sanchez, L.; Hierro, M.P.; Perez, F.J.

    2008-01-01

    The ferritic steels used to produce structural components for steam turbines are susceptible to strong corrosion and creep damage due to the extreme working conditions pushed to increase the process efficiency and to reduce pollutants release. The response of aluminide coatings on the P-92 ferritic steel, deposited by CVD-FBR, during oxidation in a simulated steam environment was studied. The analyses were performed at 650 deg. C in order to simulate the working conditions of a steam turbine, and 800 deg. C in order to produce a critical accelerated oxidation test. The Thermo-Calc software was used to predict the different solid phases that could be generated during the oxidation process, in both, coated and uncoated samples. In order to validate the thermodynamic results, the oxides scales produced during steam tests were characterized by different techniques such as XRD, SEM and EDS. The preliminary results obtained are discussed in the present work

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

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

    International Nuclear Information System (INIS)

    Ahn, Joonhong; Ikegami, Tetsuo

    2005-01-01

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

  9. Aluminum and aluminum/silicon coatings on ferritic steels by CVD-FBR technology

    International Nuclear Information System (INIS)

    Perez, F.J.; Hierro, M.P.; Trilleros, J.A.; Carpintero, M.C.; Sanchez, L.; Bolivar, F.J.

    2006-01-01

    The use of chemical vapor deposition by fluidized bed reactors (CVD-FBR) offers some advantages in comparison to other coating techniques such as pack cementation, because it allows coating deposition at lower temperatures than pack cementation and at atmospheric pressure without affecting the mechanical properties of material due to heat treatments of the bulk during coating process. Aluminum and aluminum/silicon coatings have been obtained on two different ferritics steels (P-91 and P-92). The coatings were analyzed using several techniques like SEM/EDX and XRD. The results indicated that both coatings were form by Fe 2 Al 5 intermetallic compound, and in the co-deposition the Si was incorporated to the Fe 2 Al 5 structure in small amounts

  10. Equipment installation structure of roof slab for tank type FBR and method of equipment installation

    International Nuclear Information System (INIS)

    Sakai, Takao; Yamakawa, Masanori; Otsuka, Masaya; Sekine, Katsuhisa

    1986-01-01

    Purpose: To reduce equipment thermal stress and deformation by eliminating uneven temperature distribution caused at the equipment through section of the roof slab for the tank FBR, and at the same time, simplify the structure installation. Method: Multiple number of vertical fin projects are fit on the equipment through-section inside wall for the roof slab and the cylindrical equipment peripheral wall, and with these projected fins, the ring space of the through section is vertically divided into multiple sections in the circumferential direction. The vertical fins on the through-section inside wall and the fins on the equipment peripheral wall are contacted with each other by revolving them in the lateral direction. As a result, the natural convection caused by the difference of temperatures in the vertical direction of the ring space becomes a convection within each sector divided, and never generates circumferential circulation, which reduce uneven temperature distribution caused at the equipment through section. (Kawakami, Y.)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. Application of routine methods for the inspector fuel burn-up determination and identification of displacement of spent fuel elements by dummy elements

    International Nuclear Information System (INIS)

    Rohar, S.

    1979-08-01

    14 irradiated assemblies were analyzed using nondestructive high resolution gamma spectrometry (HRGS). Measured and calculated (on the basis of calorimetric data) axial burnup profiles and average burnup values were compared. The measurements of spent fuel were performed in the Bohunice A-1 dry hot cell by using a proper collimating system and the standard Agency equipment, consisting of PGT intrinsic Ge detectors and Silena MCA with 1024 channels. The method of 134 Cs/ 137 Cs fission product activity ratio was used for burnup determination. It was found that the burnup values for 14 measured assemblies determined by HRGS were systematically lower than the calculated values with about 4-5%. The difference between the nondestructively determined burnup value of the 2N0053 assembly (average over 11 measured points) and destructively determined burnup (average over 19 measured points) was less than 2%. Passive neutron measurements of the irradiated assembly showed that the neutron counting rate was high enough for practical use and that the neutron and gamma profiles were similar and close to the burnup profile. Some calculations of gamma ray activity angular distribution were made for different numbers of dummy elements inside the irradiated assemblies. The results show that, by using gamma spectrometry transversal method, it is possible to find a significant number of dummy elements in different types of assemblies

  14. The effect of fuel micro-structure and burn-up on FGR and PCMI studied in IFA-534.13

    International Nuclear Information System (INIS)

    Matsson, I.; Teshima, H.

    1998-02-01

    Fission gas pressure (FGR) and cladding elongation (PCMI) data of four high burnup PWR fuel rods with different grain size (8.5 and 22.1 μm) have been analysed and compared in the IFA-534.13 experiment. The fission gas release is low for both fuel types. During the first part of the irradiation there is no significant difference between the normal grain size fuel and the large grain size fuel. During the second part of the experiment , the FGR appears to be higher in the large grain size fuel. However, this result should be taken with some reservation since the bellows pressure transducer showed signs of irregular behaviour during this period. The FGR at end-of-life in the large grain size fuel is #approx=#2.1 %. The FGR at end-of-life in the normal grain size fuel is #approx=#1.5 %. The degree of PCMI is higher in the large grain size fuel during the first part of the irradiation. During the second period the difference is very small. The point of interaction for PCMI during power ramps has shifted to lower power between beginning and end of irradiation. The two fuel types exhibit very similar behaviour during power ramps. There is no clear indication of relaxation during the irradiation. (author)

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

    International Nuclear Information System (INIS)

    Kobayashi, Hiroaki; Ohta, Hirokazu; Inoue, Tadashi

    2009-01-01

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

  16. Irradiation Effects in Fissile Oxides and Carbides at Low and High Burn-Up Levels; Effets de l'irradiation sur les oxydes et carbures fissiles a des taux de combustion faibles et élevés; ВЛИЯНИЕ ОБЛУЧЕНИЯ НА РАСЩЕПЛЯЩИЕСЯ ОКСИДЫ И КАРБИДЫ ПРИ НИЗКИХ И ВЫСОКИХ УРОВНЯХ ВЫГОРАНИЯ; Efectos de la irradiación sobre los óxidos y carburos fisionables para grados de combustion bajos y elevados

    Energy Technology Data Exchange (ETDEWEB)

    Frost, B. R.T.; Bradbury, B. T.; Griffiths, L. B. [Metallurgy Division, Atomic Energy Research Establishment, Harwell, Berks. (United Kingdom)

    1963-08-15

    Structural changes occurring in UO{sub 2} and in UC during irradiation to burn-ups between 100 and 10 000 MWd/t have been studied by optical and electron microscope techniques. In particular, the change in mode of fracture from transgranular to intergranular and a decrease in bond strength with increasing radiation dose have been observed. Radiation damage in UO{sub 2} has been studied by measurements of-the variation in electrical resistivity with dose on cold irradiated materials, this technique being used subsequently to study the temperature, the mechanism and the kinetics of annealing. A series of pellets of UO{sub 2}, UO{sub 2}-PuO{sub 2} solid solution, UC and PuC-UC solid solution have been irradiated in stainless-steel containers to burn-ups in the range 15000-54 000 MWd/t, the oxide centre temperatures being in the range 900-1600°C and the carbide centre temperatures about 800°C. Post-irradiation examination includes fission-gas analysis, measurement of dimensional changes and a study of structural changes by optical and electron microscopy. (author) [French] Les auteurs ont étudié au microscope optique et au microscope électronique les modifications structruelles qui se produisent dans l'UO{sub 2} et 1'UC irradiés avec des taux de combustion compris entre 100 et 10000 MWj/t. Ils ont constaté notamment un changement de mode de fracture qui, de transgranulaire, devient intergranulaire, et une diminution de la résistance au flambage lorsqu'on augmente la dose de rayonnements. Ils ont étudié les dommages causés par les rayonnements dans l'UC en mesurant les variations de la résistivité selon la dose dans des matières irradiées â froid; ils emploient ensuite cette méthode pour étudier les températures, le mécanisme et la cinétique du recuit. Une série de pastilles d'UO{sub 2}, d'UO{sub 2}-PuO{sub 2} en solution solide, d'UC et de PuC-UC en solution solide, ont été irradiées successivement dans des récipients en acier inoxydable

  17. Burn up physics; Physique des combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

    Tretiakoff, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The present communication is devoted to a body of theoretical and experimental work carried out at the C.E.A. with the aim of adding to the current knowledge on the evolution of the reactivity (during fuel irradiation) in natural or slightly enriched Uranium reactors. The difficulties of performing direct experiments on large amounts of irradiated fuels are reviewed - especially in operating power reactors - and the necessity is underlined for fundamental research in two directions: on one hand, the change in the composition of the fuels (chains of heavy nuclei, fission products), and on the other hand the effect of changes in composition on the neutron balance. Before presenting three types of experiments which have been carried out, the importance of the problems associated with the neutron spectra is stressed and the practical methods used for the calculations are briefly described. The systematic irradiation of several types of fuel, followed by their chemical and isotopic analysis has been going on for several years. An outline of the experimental programme is given with a description of the methods employed: {alpha}, {beta}, {gamma} chain for the preparation of samples determination of the plutonium content by coulometry and double isotopic dilution, separation of Boron used in some cases for the measurement of integrated neutron densities. The interpretation of the measurements is discussed with some examples. A second and more recent series of experiments deals with the investigation of lattices, using synthetic fuels (Uranium-Plutonium alloys) as compared to slightly depleted or enriched Uranium Various experiments are considered on heavy water and on cold graphite, then on graphite heated up to 500 C Some results already obtained are listed. These experiments, requiring nearly a metric ton of each type of fuel cannot be pursued in a systematic manner. This is why is developed since several years a method of differential measurement by oscillation, which requires samples of the order of several kilograms only. The relationships between these measurements and the investigations of lattices are discussed, and an outline is given of the way of carrying out the systematic study of fuels of various compositions. The method has been successfully applied to the systematic study of irradiated fuels (analysed independently by the methods mentioned above) thus giving the possibility of measuring in situ the absorption of fission products. (author) [French] Cette communication expose un ensemble d'etudes theoriques et d'experiences, effectuees au CEA et destinees a faire progresser la connaissance de l'evolution de la reactivite (au cours de l'irradiation du combustible) dans les reacteurs a uranium naturel ou faiblement enrichi.,. On rappelle les difficultes de l'experimentation directe sur des masses importantes de combustible irradie - en particulier dans les reacteurs de puissance en fonctionnement - et on souligne la necessite d'experiences a caractere fondamental distinguant: d'une part l'evolution de la composition des combustibles (chaines de noyaux lourds, produits de fission), d'autre part l'effet des modifications de composition sur le bilan de neutrons. Avant de presenter trois categories d'experiences que l'on est conduit a entreprendre, on rappelle l'importance des problemes lies aux spectres de neutrons et on decrit rapidement les methodes pratiques de calcul utilisees. L'irradiation systematique de quelques types de combustibles, suivie de leur analyse chimique et isotopique est en cours depuis plusieurs annees. On donne un apercu de l'ensemble du programme experimental et on decrit les moyens et les methodes mis en oeuvre: chaine {alpha}, {beta}, {gamma} pour la preparation des echantillons, dosage du Plutonium par coulommetrie et double dilution isotopique, separation du Bore utilise dans certains cas pour la mesure des densites de neutrons integrees. On discute sur quelques exemples l'interpretation des mesures. Une deuxieme serie d'experiences, plus recente, porte sur l'etude de reseaux utilisant des combustibles reconstitues (alliages Uranium - Plutonium), compares a de l'Uranium legerement appauvri ou enrichi. On decrit les experiences prevues (eau lourde, graphite a froid puis jusqu'a 500 C) et on expose les resultats deja obtenus. Ces experiences, necessitant pres d'une tonne de combustible de chaque espece, ne peuvent se preter a des etudes systematiques. C'est pourquoi on a poursuivi depuis quelques annees la mise au point d'une methode de mesure differentielle par oscillation, ne necessitant que des echantillons de l'ordre de quelques kilogrammes. On discute les relations entre ces mesures et les etudes de reseaux et on montre comment est effectuee l'etude systematique de compositions de combustible variees. La methode a pu etre appliquee avec succes a l'etude systematique de combustibles irradies (soumis par ailleurs aux analyses citees dans la premiere partie) donnant ainsi la possibilite de mesurer in situ l'absorption des produits de fission. (auteur)

  18. Mineralization and defluoridation of 2,2,3,3-tetrafluoro -1-propanol (TFP) by UV oxidation in a novel three-phase fluidized bed reactor (3P-FBR).

    Science.gov (United States)

    Shih, Yu-Jen; Tsai, Meng-Tso; Huang, Yao-Hui

    2013-05-01

    2,2,3,3-Tetrafluoro-1-propanol (TFP, C3H4F4O, M.W. = 132.06) is extensively used as the solvent in CD-R and DVD-R fabrication. Since it has a fluorinated alky-chain configuration and is non-biodegradable, its treatment by conventional oxidation methods is typically very inefficient. In this work, novel three-phase fluidized bed reactor (3P-FBR, 7.5 cm in diameter, 50 cm high) that combines photo oxidation (UV/H2O2, one of AOPs (Advanced Oxidation Process) and adsorption (BT5 iron oxide as adsorbent) processes is designed for mineralizing and defluorinizing TFP wastewater. The experimental results reveal that TFP can be efficiently mineralized, and the BT5 that is circulated by aeration in the 3P-FBR system can remove the released fluoride ions in the reaction period. Irradiation with 254 nm UV and a 10 mM H2O2 dose yield a TOC removal of TFP (1.39 mM, equivalent to an initial TOC of 50 ppm) of over 99.95% in 2 h, and 99% of fluoride was removed by BT5 with an adsorption capacity of 24.1 mg-F g(-1). Copyright © 2013 Elsevier Ltd. All rights reserved.

  19. A basic research on the transient behavior for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

    1999-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  1. Investigations of chemical reactions between U-Zr alloy and FBR cladding materials

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Ukai, Shigeharu

    2005-07-01

    U-Pu-Zr alloys are candidate materials for commercial FBR fuel. However, informations about chemical reactions with cladding materials developed by JNC for commercial FBR have not been well obtained. In this work, the reaction zones formed in four diffusion couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS, U-10wt.%Zr/12Cr-ODS, and U-10wt.%Zr/Fe at about 1013K have been examined and following results were obtained. 1) At about 1013K, in the U-10wt.%Zr/Fe couple, the liquid phase zones were obtained. In the other couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS and U-10wt.%Zr/12Cr-ODS, no liquid phase zones were obtained. The obtained chemical reaction zones in the later 3 couples were similar to the reported ones obtained in U-Zr/Fe couples without liquid phase formation. In comparison with the reaction zones obtained in the U-10wt.%Zr/Fe couple, the reaction zones inside cladding materials obtained in the PNC-FMS, 9Cr-ODS, and 12Cr-ODS couples were thin. 2) From the investigations of relationship between the obtained depths of the chemical reaction zones inside cladding materials and composition of the cladding materials, it was considered that the depth of chemical reaction zone would depend on the Cr content of the cladding materials and the depth would decrease with increasing Cr content, resulting in prevention of liquid phase formation. 3) From the investigations of the mechanisms of chemical reactions between U-Pu-Zr/cladding materials, it was considered that the same effect of Cr obtained in the U-Zr/cladding materials would be expected in U-Pu-Zr/cladding materials. Those seemed to indicate that the threshold temperatures of liquid phase formation for U-Pu-Zr/PNC-FMS, U-Pu-Zr/9Cr-ODS, and U-Pu-Zr/12Cr-ODS might be higher than that for U-Pu-Zr/Fe. (author)

  2. Electronic manual of the nuclear characteristics analysis code-set for FBR

    International Nuclear Information System (INIS)

    Makino, Tohru

    2001-03-01

    Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

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

    International Nuclear Information System (INIS)

    Zhou Yingxue; Wang Dongsheng; Zhang Xinyi

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Nakagawa, Masaki; Shigeta, Masayuki.

    1984-01-01

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

  5. Method of start-up of rotary plug sealing devices in FBR type reactors

    International Nuclear Information System (INIS)

    Sakuragi, Masanori; Akita, Haruo

    1980-01-01

    Purpose: To rapidly and safely start-up the rotary plug sealing device by controlling to eliminate the pressure difference in the pressures of gases exerting on the liquid surfaces in the inner and the outer cylinders of a sealing alloy vessel in the rotary plug of a FBR type reactor. Method: In a case where an abnormal state results in the pressure difference of gases exerted on the liquid surfaces in the inner and the outer cylinders of a vessel charged with sealing alloy in a rotary plug and the sealing valve for the back-up gas supply tube is rapidly closed to seal the sealing portion, the pressure in the gas supply tube is controlled so that the pressure difference in the gases exerted on the liquid surfaces in the inner and outer cylinders while closing the sealing valve. Then, after conforming that the pressure is controlled to a predetermined level at which the pressure difference can be regarded to be zero, the sealing valve is gradually opened while regulating the pressure in the gas supply tube so as to maintain the pressure difference to a predetermined level. This prevents the occurrence of external disturbances upon opening of the sealing valve and enables rapid and safety start-up for the rotary plug sealing device. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

    Ueda, Nobuyuki; Kinoshita, Izumi; Nishi, Yoshihisa

    2000-01-01

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

  7. High order depletion sensitivity analysis

    International Nuclear Information System (INIS)

    Naguib, K.; Adib, M.; Morcos, H.N.

    2002-01-01

    A high order depletion sensitivity method was applied to calculate the sensitivities of build-up of actinides in the irradiated fuel due to cross-section uncertainties. An iteration method based on Taylor series expansion was applied to construct stationary principle, from which all orders of perturbations were calculated. The irradiated EK-10 and MTR-20 fuels at their maximum burn-up of 25% and 65% respectively were considered for sensitivity analysis. The results of calculation show that, in case of EK-10 fuel (low burn-up), the first order sensitivity was found to be enough to perform an accuracy of 1%. While in case of MTR-20 (high burn-up) the fifth order was found to provide 3% accuracy. A computer code SENS was developed to provide the required calculations

  8. Measurement of burnup in FBR MOX fuel irradiated to high burnup

    International Nuclear Information System (INIS)

    Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

    2003-01-01

    The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

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

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi

    2001-11-01

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

  10. Development of disassembly and pin chopping technology for FBR spent fuels

    International Nuclear Information System (INIS)

    Kobayashi, Tsuguyuki; Namba, Takashi; Kawabe, Yukinari; Washiya, Tadahiro

    2008-01-01

    Japan Atomic Power Company (JAPC) and Japan Atomic Energy Agency (JAEA) have been developing fuel disassembly and fuel pin chopping systems for a future Japanese commercial FBR. At first, the wrapper tube is cut by the slit-cut to pull it out, then the fuel pins are cut by the crop-cut at their end-plugs to separate them from the entrance nozzle. The pins are transferred to the magazine of the chopping machine. A series of tests were performed to develop this procedure. As the result of mechanical cutting tests, the CBN wheel was selected. The slit-cut tests were carried out to evaluated the cutting performance of the wheel. The wrapper tube is normally slit-cut in the circumferential direction. One CBN wheel could cut more than 5 fuel assemblies in this direction. The slit-cut in the axial direction is prepared as provision when the tube is difficult to put out. More work is needed to cut 5mm thick PNC-FMS plate in this direction without damaging the pins beneath it. As the result of the crop-cut tests of end-plugs made of ODS steel, the CBN wheel could cut the 61 pin bundle by two strokes. More work is needed to cut the 217 pin bundle. Fuel pin handling tests were performed to transfer them from the disassembly machine to the chopping machine. The Saucer tray was selected to receive the disassembled pins. All the pins were transferred and loaded into a magazine of the chopping machine. Fuel pin loading tests were conducted to optimize the magazine configuration to make the chopping length within 1.0±0.5 cm. In order to decrease the disturbance during chopping, the width of the magazine was adjusted to be 12 cm and installation of a height adjuster is favourable to control the free space above the pins. (author)

  11. Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

    International Nuclear Information System (INIS)

    Goswami, S.

    1983-01-01

    The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

  12. 3-D numerical simulation on the vibration of liquid sodium's free surface in sodium pool of FBR

    International Nuclear Information System (INIS)

    Han Biao; Yao Zhaohui; Ye Hongkai; Wang Xuefang

    1997-01-01

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

  13. MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

    International Nuclear Information System (INIS)

    Sugino, K.; Iwai, T.

    2006-01-01

    MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

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

    International Nuclear Information System (INIS)

    Ohkawachi, Yasushi; Fukushima, Manabu

    1999-05-01

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

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

    International Nuclear Information System (INIS)

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

    1998-10-01

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

  16. The basic research on the CDA initiation phase for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime

    1998-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

  17. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

  18. Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties. (author)

  19. Non-Destructive Methods for Determining Burn-Up in Nuclear Fuel; Methodes Non Destructives d'Evaluation du Taux de Combustion dans le Combustible Nucleaire; Metody opredeleniya vygoraniya v yadernom toplive bez razrusheniya obraztsa; Metodos No Destructivos para Determinai el Grado de Combustion de los Elementos Combustibles Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    McGonnagle, W. J. [Illinois Institute of Technology, Chicago, IL (United States)

    1966-02-15

    Non-destructive methods for quantitative measurement of burn-up in nuclear reactor fuel elements are useful and desirable. The ideal method for fuel assay would be one that requires no special information about the neutron spectra, radiation history, or cooling time. The irradiated fuel element contains a record of the fuel burn-up. This record is in the form of radioactive and stable isotopes resulting from the fission process. Unfortunately, in the non-destructive as well as the destructive fuel assay methods, the neutron spectrum, irradiation history, and cooling period influence this record. Likewise, the lack of precise nuclear data, such as values of nuclear cross-sections, affects any calculations that can be made. Another difficulty in the non-destructive assay is the presence of high radiation fields which contribute to the ''noise'' background of the measurements. The development of useful and realistic standards is difficult. The non-destructive burn-up methods do serve a useful purpose especially when an approximate value of burn-up is required quickly and economically even though in the present state of the art they lack the desired precision and accuracy. Several non-destructive methods for determining burn-up have been used, are being evaluated, or have been proposed. Various types of spectrometers including the bent crystal, magnetic Compton, Compton coincidence, and scintillation have been used to analyse the gamma radiation from the radioactive material formed during the fission process. Other non-destructive methods include foil activation, neutron transmission, activation analysis, measurement of capture gamma rays, and the measurement of prompt and delayed neutrons. The basic principles of each of the above instruments and methods, their sensitivities and their limitations will be reviewed. Non-destructive methods using stable isotopes produced during the fission process are proposed. In the use of stable isotopes, detailed irradiation history

  20. Assessment of the linear power level in fuel rods irradiated in the CALLISTO loop in the high flux materials testing reactor BR2

    International Nuclear Information System (INIS)

    Malambu, E.; Raedt, Ch. de; Weber, M.

    1999-01-01

    The pressurized light-water-cooled testing facility CALLISTO was designed to test the behaviour of advanced fuel rods (UO 2 or MOX, possibly with burnable poisons) under conditions representative of actual LWRs up to high burn-up rates. The accurate determination of the fission powers in each of the nine rods, and hence of the burn-up values, is carried out according to a rather elaborate procedure. (author)

  1. Standardization of dosimetry and damage analysis work for U.S. LWR, FBR, and MFR development program

    International Nuclear Information System (INIS)

    McElroy, W.N.; Doran, D.G.; Gold, R.; Morgan, W.C.; Grundl, J.A.; McGarry, E.D.; Kam, F.B.K.; Swank, J.H.; Odette, G.R.

    1978-01-01

    The accuracy requirements for various measured/calculated exposure and correlation parameters associated with current dosimetry and damage analysis procedures and practices depend on the accuracy needs of reactor development efforts in testing, design, safety, operations, and surveillance programs. Present state-of-the-art accuracies are estimated to be in the range of +-2 to 30 percent (1 sigma), depending on the particular parameter. There now appears to be international agreement, at least for the long term, that most reactor fuels and materials programs will not be able to accept an uncertainty greater than about +5 percent (1 sigma). The current status of dosimetry and damage analysis standardization work within the U.S. for LWR, FBR and MFR is reviewed in this paper

  2. Studies on the seismic buckling design guideline of FBR main vessels. 9. Buckling evaluation under elastic-plastic seismic response

    International Nuclear Information System (INIS)

    Hagiwara, Yutaka; Yamamoto, Kohsuke; Kawamoto, Yoji; Nakagawa, Masaki; Akiyama, Hiroshi

    1998-01-01

    Plastic shear-bending buckling under seismic loadings is one of the major problems in the structural design of FBR main vessels. Pseudo-dynamic and dynamic buckling tests of cylinders were performed in order to study the effects of nonlinear seismic response on buckling strength, ductility, and plastic response reduction. The buckling strength formulae and the rule for ductility factors both derived from static tests were confirmed to be valid for the tests under dynamic loads. The displacement-constant rule for response reduction effect was modified by acceleration amplification factor in order to maintain applicability for various spectral profiles of seismic excitations. The response reduction estimated by the proposed rule was reasonably conservative for all cases of the pseudo-dynamic and the dynamic tests. Finally, a seismic safety assessment rule was proposed for plastic shear-bending buckling of cylinders, which include the proposed response reduction rule. (author)

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

    International Nuclear Information System (INIS)

    Mihalache, Ovidiu

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Kato, Shoichi; Yoshida, Eiichi

    2002-12-01

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

  5. Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burnup condition

    International Nuclear Information System (INIS)

    Kim, H. D.; Kim, D. H.; Park, S. Y.; Park, J. H.

    2005-10-01

    This project was performed by KAERI in the frame of construction of the international cooperative basis on the nuclear energy. This was supported from MOST under the title of 'Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burn up condition'. The current operating NPP are converting the burned fuel to the wasted fuel after burn up of 40 GWD/MTU. But in Korea, burn up of more than 60 GWD/MTU will be expected because of the high fuel efficiency but also cost saving for storing the wasted fuel safely. The domestic research for the purpose of developing the fuel and the cladding that can be used under the high burn up condition up to 100 GWD/MTU is in progress now. But the current computer code adopts the model and the data that are valid only up to the 40 GWD/MTU at most. Therefore the current model could not take into account the phenomena that may cause differences in the fission product release behavior or in the core damage process due to the high burn up operation (more than 40 GWD/MTU). To evaluate the safety of the NPP with the high burn up fuel, the improvement of current severe accident code against the high burn up condition is an important research item. Also it should start without any delay. Therefore, in this study, an expert group was constructed to establish the research basis for the severe accident under high burn up conditions. From this expert group, the research items regarding the high burn up condition were selected and identified through discussion and technical seminars. Based on these selected items, the meeting between IRSN and KAERI to find out the cooperative research items on the severe accident under the high burn up condition was held in the IRSN headquater in Paris. After the meeting, KAERI and IRSN agreed to cooperate with each other on the selected items, and to co-host the international seminar, and to develop the model and to

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  7. Post-irradiation examinations and high-temperature tests on undoped large-grain UO{sub 2} discs

    Energy Technology Data Exchange (ETDEWEB)

    Noirot, J., E-mail: jean.noirot@cea.fr [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Pontillon, Y. [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Yagnik, S. [EPRI, P.O. Box 10412, Palo Alto, CA 94303-0813 (United States); Turnbull, J.A. [Independent Consultant (United Kingdom)

    2015-07-15

    Within the Nuclear Fuel Industry Research (NFIR) programme, several fuel variants –in the form of thin circular discs – were irradiated in the Halden Boiling Water Reactor (HBWR) at burn-ups up to ∼100 GWd/t{sub HM}. The design of the fuel assembly was similar to that used in other HBWR programmes: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature differences within each fuel disc. One such variant was made of large-grain UO{sub 2} discs (3D grain size = ∼45 μm) which were subjected to three burn-ups: 42, 72 and 96 GWd/t{sub HM}. Detailed characterizations of some of these irradiated large-grain UO{sub 2} discs were performed in the CEA Cadarache LECA-STAR hot laboratory. The techniques used included electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS). Comparisons were then carried out with more standard grain size UO{sub 2} discs irradiated under the same conditions. Examination of the high burn-up large-grain UO{sub 2} discs revealed the limited formation of a high burn-up structure (HBS) when compared with the standard-grain UO{sub 2} discs at similar burn-up. High burn-up discs were submitted to temperature transients up to 1200 °C in the heating test device called Merarg at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during these tests, the release peaks throughout the temperature ramp were monitored. Tests at 1600 °C were also conducted on the 42 GWd/t{sub HM} discs. The fuels were then characterized with the same microanalysis techniques as those used before the tests, to investigate the effects of these tests on the fuel’s microstructure and on the fission gas behaviour. This paper outlines the high resistance of this fuel to gas precipitation at high temperature and to HBS formation at high burn-up. It also shows the similarity of the positions, within the grains, where HBS forms

  8. FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Sakagami, M.

    1984-01-01

    FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

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

    International Nuclear Information System (INIS)

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

    1996-06-01

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

  11. Feasibility study on ductility exhaustion approach for creep-fatigue damage assessment of FBR 316 stainless steel using published data

    International Nuclear Information System (INIS)

    Nonaka, Isamu; Kitagawa, Masaki; Torihata, Shoji.

    1995-01-01

    In order to investigate the applicability of a ductility exhaustion rule to the creep-fatigue life assessment of FBR 316 stainless steel, a feasibility study using the published data was conducted. The assessment method was proposed based on the linear damage summation rule. In the proposed method, fatigue damage was calculated by Minor's rule and creep damage was calculated by a ductility exhaustion rule. The creep-fatigue lives in the published data were predicted by the proposed method. The results obtained are as follows: (1) All the data could be predicted within a factor of two on life by the proposed method. (2) The creep-fatigue lives under 10 minute strain hold at 550degC were overestimated, while those under 60 minute strain hold at 550degC and 600degC were estimated adequately. From the above facts, the proposed method seemed to be effective for the prediction of creep-fatigue life in which the creep damage was dominant and also the intergranular cracking was remarkable. (3) The creep damage was simultaneously calculated by the time fraction rule in order to compare with the ductility exhaustion role. All the data could be also predicted within a factor of two on life by this rule, but it tended to overestimate the life. (author)

  12. Development of a FBR fuel pin bundle deformation analysis code 'BAMBOO' . Development of a dispersion model and its validation

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ukai, Shigeharu; Asaga, Takeo

    2002-03-01

    Bundle Duct Interaction (BDI) is one of the life limiting factors of a FBR fuel subassembly. Under the BDI condition, the fuel pin dispersion would occur mainly by the deviation of the wire position due to the irradiation. In this study the effect of the dispersion on the bundle deformation was evaluated by using the BAMBOO code and following results were obtained. (1) A new contact analysis model was introduced in BAMBOO code. This model considers the contact condition at the axial position other than the nodal point of the beam element that composes the fuel pin. This improvement made it possible in the bundle deformation analysis to cause fuel pin dispersion due to the deviations of the wire position. (2) This model was validated with the results of the out-of-pile compression test with the wire deviation. The calculated pin-to-duct and pin-to-pin clearances with the dispersion model almost agreed with the test results. Therefore it was confirmed that the BAMBOO code reasonably predicts the bundle deformation with the dispersion. (3) In the dispersion bundle the pin-to-pin clearances widely scattered. And the minimum pin-to-duct clearance increased or decreased depending on the dispersion condition compared to the no-dispersion bundle. This result suggests the possibility that the considerable dispersion would affect the thermal integrity of the bundle. (author)

  13. High level nuclear wastes

    International Nuclear Information System (INIS)

    Lopez Perez, B.

    1987-01-01

    The transformations involved in the nuclear fuels during the burn-up at the power nuclear reactors for burn-up levels of 33.000 MWd/th are considered. Graphs and data on the radioactivity variation with the cooling time and heat power of the irradiated fuel are presented. Likewise, the cycle of the fuel in light water reactors is presented and the alternatives for the nuclear waste management are discussed. A brief description of the management of the spent fuel as a high level nuclear waste is shown, explaining the reprocessing and giving data about the fission products and their radioactivities, which must be considered on the vitrification processes. On the final storage of the nuclear waste into depth geological burials, both alternatives are coincident. The countries supporting the reprocessing are indicated and the Spanish programm defined in the Plan Energetico Nacional (PEN) is shortly reviewed. (author) 8 figs., 4 tabs

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

    International Nuclear Information System (INIS)

    Hamada, H.; Kurihara, A.

    2003-05-01

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

  15. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  16. Evaluation of very low frequencies of ATWS and PLOHS in a loop-type FBR plant by making use of inherently safe features

    International Nuclear Information System (INIS)

    Sakata, K.; Koyama, K.; Aoi, S.; Simonelli, R.B.; Wallace, I.T.

    1987-01-01

    Frequencies of ATWS (Anticipated Transient Without Scram) and PLOHS (Protected Loss of Heat Sink) for a large loop-type FBR plant were evaluated by applying PSA methodologies. The frequencies were found to be so low that ATWS and PLOHS could be excluded from candidates of the design basis events. Furthermore, the inherently safe features introduced to the system design were verified to be very effective for reduction of the Probability of CCF (Common Cause Failure), which deteriorates reliability of both the reactor shutdown and the decay heat removal systems. (orig.)

  17. Summary remarks and recommended reactions for an international data file for dosimetry applications for LWR, FBR, and MFR reactor research, development and testing programs

    International Nuclear Information System (INIS)

    McElroy, W.N.; Lippincott, E.P.; Grundl, J.A.; Fabry, A.; Dierckx, R.; Farinelli, U.

    1979-01-01

    The need for the use of an internationally accepted data file for dosimetry applications for light water reactor (LWR), fast breeder reactor (FBR), and magnetic fusion reactor (MFR) research, development, and testing programs continues to exist for the Nuclear Industry. The work of this IAEA meeting, therefore, will be another important step in achieving consensus agreement on an internationally recommended file and its purpose, content, structure, selected reactions, and associated uncertainy files. Summary remarks and a listing of recommended reactions for consideration in the formulation of an ''International Data File for Dosimetry Applications'' are presented in subsequent sections of this report

  18. Research and development on materials, structural strength and seismic integrity of FBR components

    International Nuclear Information System (INIS)

    Sumikawa, Masaharu; Kirihara, Seishin; Shigeta, Masayuki; Shimoyashiki, Shigehiro; Nishioka, Akio.

    1982-01-01

    For designing high temperature structures of FBRs, highly reliable design is required on the basis of safety requirement. At the same time, it is necessary to guarantee the soundness of structures over the total design life of plants. Since the high temperature equipments are operated in a creep temperature region and show nonlinear behaviour, nonlinear structural analysis is required. Hitachi Ltd., based on the concept of verifying the latest technology to reflect it to the design along with its adoption, has progressed various research and development by organizing a project team collecting specialists in the company, independently developing and modifying the nonlinear structural analysis and evaluation program, and establishing the organization through the introduction of a general purpose large scale computer. The research and development for materials include the development of the strength standards for high temperature structural materials and the improvement of the high temperature characteristics of JIS stainless steel SUS 321. In the R and D for high temperature strength, the test on the deforming behaviour of plates due to bending creep, the thermal shock test for steam generator tube plates and others were performed. In the R and D for seismic integrity, the vibration test of piping support structure and the development of detailed seismic property evaluation program are mentioned. (Wakatsuki, Y.)

  19. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

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

    International Nuclear Information System (INIS)

    Kawasaki, Nobuchika; Asayama, Tai

    2001-09-01

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

  1. Study on the FBR cycle introduction scenario. 2. A study on the role of nuclear energy under the diversity of energy supply-and-demand

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ono, Kiyoshi; Hirao, Kazunori

    2002-03-01

    This report concerns it self with the results of an investigation about the possibility of future nuclear utilization in the part of FBR Cycle Introduction Scenario Study in the JNC's 'Feasibility Study on Commercialized Fast Reactor Cycle System (the F/S)'. We have investigated about the problems that confront energy industries and electric power companies, the capacities of distributed generation, the coexistence method of a distributed generation and large-scale power supply generation, and the development status of a small-scale nuclear reactor from a wide viewpoint. Especially the spread of distributed generation causes the decrease of the electricity demand which the electric power companies supplies. Since introduction scale of a distributed power supply is also expected to increase in the future, it will give some influences to a future nuclear plan and a power supply plan. The hydrogen utilization with out greenhouse gas mission is expected to spread with distributed generation, such as a fuel cell and a micro-gas turbine. Therefore, we proposed the new business model that the hydrogen produced by using nuclear surplus electricity is consumed distributed generation, such as a fuel cell and a micro-gas turbine. We plan to evaluate quantitatively the best power supply composition based on this load stability business model, FBR introduction capacities, the load factor, and the amount of CO 2 reduction. (author)

  2. Research and development on in-service inspection system for reactor vessel of FBR's

    International Nuclear Information System (INIS)

    Rindo, Hiroshi; Mitabe, Noriaki; Ara, Kuniaki; Nagai, Keiichi; Otaka, Masahiko

    1993-01-01

    In-Service Inspection (ISI) is required for main components and piping of FBRs. Visual test and volumetric examination of the reactor vessel (RV) from the outer surface are to be performed under severe conditions such as limited space, high temperature and high gamma dose rate during the reactor shutdown. Therefore, ISI should be performed by using a remote operation system, and the ISI system should be very compact. PNC has been developing the ISI system to apply to the RV inspection. Verification and performance tests of ISI system were carried out by use of the RV test model. This paper describes the system structure, system verification tests including operation and controlling the inspection robot, the functions of the visual test and the volumetric examination under the high temperature

  3. Research on creation of new materials for innovative improvement of FBR performance

    International Nuclear Information System (INIS)

    Kano, S.; Yoshida, E.; Inoue, M.

    1992-01-01

    Creation of new materials is an essential issue to attain an innovative improvement of fast reactors' performance to achieve its excellent economics. In this paper, status and highlights are presented on the research of new materials in PNC. Firstly, sodium corrosion characteristics has been investigated on new ceramics recently developed and some considerations for their improvement have been clarified to create advanced ceramics having an excellent resistance to sodium corrosion. Secondary, materials design and manufacturing process by powder metallurgy were investigated to create new composites for radiation shielding. High performance shielding materials having an excellent shielding characteristics and a high heat resistance have been created. Thirdly, status of investigation on materials design and manufacturing process is presented on functionally gradient materials to create high performance thermal stress relieving material and innovative core material. Fourthly, manufacturing process by powder metallurgy was investigated on new composites such as B 4 C cermets, etc., to create high performance fission reaction control materials. Trial production pellets were evaluated on micro-structure, thermal conductivity, anti-thermal shock properties. etc.. (author)

  4. Ultrasonic inspection of steam generator tubes in Superphenix F.B.R. Power plant

    International Nuclear Information System (INIS)

    Gondard, C.

    1991-01-01

    An ultrasonic method has been developed to test of the S.G's tubes of SPX fast breeder reactor. A new type of rotating probes for cracks and wall thickness measurements have been built up and successfully tested. The data acquisition and processing system SPARTACUS was used; it allows high frequency digitalization and powerful signal processings using frequency representations. The actual performances were tested on natural defects under representative operating conditions

  5. USA/FBR program fast flux test facility startup physics and reactor characterization methods and results

    International Nuclear Information System (INIS)

    Bennett, R.A.; Harris, R.A.; Daughtry, J.W.

    1981-09-01

    Final confirmation of much of the engineering mockup work has been achieved in FTR zero-power experiments in February, 1980, and in power demonstration performed in December, 1980, and March, 1981. Final in-core low-power and high-power irradiation of spatially distributed radioactivants will be completed late in 1981. This paper describes physics experiments and present summaries of the extensive results accumulated to date. 53 figures

  6. Analysis of fluid-structure interaction mechanism of a Na-FBR core while the evacuation of a gas pocket

    International Nuclear Information System (INIS)

    Sargentini, Lucia

    2014-01-01

    The purpose of this study is to improve the knowledge about the core behavior of a sodium fast breeder reactor (Na-FBR) during vibrations through the fluid-structure interaction analysis. Namely, we investigate the flowering of the Phenix core during the SCRAM for negative reactivity (AURN) and the seismic behavior of the core of Astrid project. Three approaches are followed: experimental campaign, performing of analytical solution and development of numerical model. We create a flow regime map to identify the flow regimes in the fluid gap for very short times scales (as AURN) as well as longer time scales (as seismic oscillations). The most suitable equation system (Navier-Stokes, Euler or linearized Euler) is chosen to model the fluid flow in the numerical code. To our knowledge, for the first time, an analytical solution for free vibration and very narrow gaps is proposed. We designed two experimental apparatus (PISE-1a and PISE-2c) composed respectively by 1 and 19 hexagonal assemblies (two crowns) of Poly-methyl methacrylate (PMMA). Every PMMA assembly is fixed to a stainless steel twin-blades support allowing only orthogonal oscillations with respect to generating line of assembly. The twin-blades supports are designed to give the same range frequency of Phenix assembly in liquid sodium. The experimental equipment PISE-1a is used to determine the dynamic characteristics of PISE-2c assembly, to calibrate instrumentation and for validating our numerical model. Free vibration tests in air are performed to evaluate the dynamic characteristics of the body. Free vibration experiments in water allow to assess the added mass and added damping effect on the frequency. Even though the fluid flow during vibration should be completely bidimensional, the fluid flow is affected by a 3D effect - named 'jambage' - at the top and the basis of the assembly. This effect produces a lower frequency than the theoretical value. Tests are modeled with a bidimensional

  7. Evaluation of bundle duct interaction by out-of-pile compression test of FBR fuel pin bundles

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Kosuke; Yamamoto, Yuji; Nagamine, Tsuyoshi; Maeda, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2001-06-01

    Bundle duct interaction (BDI) caused by expansion of fuel pin bundle is a main factor to limit the fuel lifetime. Therefore, it is important for the design of fast reactor fuel assembly to understand the fuel pin deformation behavior under BDI condition. In order to understand the fuel pin deformation behavior under BDI condition, out-of-pile compression tests were conducted for FBR fuel pin bundle by use of X-ray CT equipment. In these compression tests, two kinds of fuel pin bundles were conducted. One was the fuel pin bundle with the short wire-pitch and the other was the fuel pin bundle with the short wire-pitch and large diameter claddings. The general discussions were also performed based on the results of out-of-pile compression tests obtained by use of X-ray CT equipment in the previous work. Following results were obtained. 1) The occurrence of the pin-to-duct contact depends on the wire-pitch. In the fuel pin bundle with large wire-pitch, the pin-to-duct contact occurred at the early stage of BDI. The reason of this result is due to the low bowing rigidity of the fuel pins with long wire-pitch. 2) The value of the ovalation stiffness strongly depends on the geometry of cladding (diameter, thickness) and especially on wire-pitch. This result in this work revealed that the occurrence of the pin-to-duct contact depends on the value of the ovalation stiffness. 3) The occurrence of wire dispersion and dispersive displacement of pins depends on the wire-pitch strongly. In the fuel pin bundle with the long wire-pitch, the occurrence of the above-mentioned suppression mechanism to BDI is remarkable. 4) The suppression mechanism to BDI of the fuel pin bundle with the long wire-pitch is elastic oval deformation of cladding, wire dispersion and dispersive displacement of pins. On the other hand, the elastic and plastic oval deformation of cladding is the major suppression mechanism to BDI in the fuel pin bundle with the short wire-pitch. 5) The appearance of

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

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

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

  9. Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

    International Nuclear Information System (INIS)

    Sonoda, Yukio; Tamaoki, Tetsuo

    1989-01-01

    Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

  10. Physics and behaviour during a ULOF of an innovative heterogeneous annular FBR core

    International Nuclear Information System (INIS)

    Massara, S.; Verwaerde, D.

    2012-01-01

    The major conclusions: • The reduction of the Na void worth is a way allowing a strong improvement of the dynamic behavior in very severe ULOF transient (10 s halving time), possibly allowing to avoid Na boiling; • 1st order effects: Na density ( 0); • 2nd order effects: - Mass flow gaggling scheme (as a function of the core neutronics); - Other feed-back effects: diagrid, driveline feed-back. → Very strong impact of uncertainties: Thermalhydraulic models & codes, drive-line feed-back modeling; • Methodology for feed-back coefficient calculation (example: in this calculation the Na density effect is linearized from nominal to 100% void, anticonservative in case of no Na boiling); • Core neutronics: nuclear data, models. → Even in case of no Na boiling, the critical events will be: • Fuel cladding and S/A wrapper behavior at very high temperature; • Upper core structures behavior

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

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi

    2002-03-01

    The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: 1. To evaluate the structural integrity of tube material, the strength standard for 2. 25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200degC) creep data. This standard has been validated with the tube rupture simulation test data. 2. The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. 3. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. 4. The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. 5. The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system. (author)

  12. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  13. A preliminary design study of a pool-type FBR 'ARES' eliminating intermediate heat transport systems

    International Nuclear Information System (INIS)

    Ueda, N.; Nishi, Y.; Kinoshita, I.; Yoshida, K.

    2001-01-01

    An innovative reactor concept 'ARES' (Advanced Reactor Eliminating Secondary system) is proposed to aim at reducing the construction cost of a liquid metal cooled fast breeder reactor (LMFBR). This concept is developed to show the ultimate cost down potential of LMFBR's at their commercial stage. The electrical output is 1500 MW, while the thermal output is 3900 MW. Main components of the primary cooling system are four electromagnetic pumps (EMP) and eight double-wall-tube steam generators (SG). Both of them are installed in a reactor vessel like pool type LMFBR's. An intermediate heat transport system which a previous LMFBR has it eliminated, main components of which are intermediate heat exchangers (IHX), secondary pumps and secondary piping. Further, a high reliable SG could decrease the occurrence of water leak accidents and reduce the related mitigation systems. In this study, structure concept, approach to embody a high reliable SG and accidents analyses are carried out. Flow path configuration is mainly discussed in investigation of the structure concept. In case of a water leak accident in a SG, the fault SG must be isolated to prevent a reaction production from flowing into the core. The measure to cut both inlet and outlet coolant flow paths by siphon-break mechanism is adopted to be consistent with the decay heat removal operation. The safety design approach of the double-wall-tube SG is investigated to limit the accident occurrence below 10 -7 (1/ry). A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to prevent adhesion of the double-wall-tube effectively. The reliability of the tube-to-tube-sheet was evaluated as 10 -10 (1/hr) for an inner tube and 10 -9 (1/hr) for an outer tube with reference to the failure experience of previous SG's. The failure must be detected within 60 to 120 minutes. Finally, a seamless U tube type of double-wall-tube SG is adopted. Transient events due to

  14. Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    International Nuclear Information System (INIS)

    Ito, Masahiro; Uwaba, Tomoyuki

    2005-04-01

    JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

  15. Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up

    Science.gov (United States)

    Faghihi, F.; Saidi-Nezhad, M.

    2009-09-01

    A set of coupled IVPs that describe the change rate of an important poison, in a nuclear reactor, has been written herein. Specifically, in this article, we have focused on the samarium-149 (as a poison) burnup in a desired pressurized water nuclear reactor and its concentration are given using our MATLAB-linked "solver."

  16. Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up

    International Nuclear Information System (INIS)

    Faghihi, F.

    2009-01-01

    A set of coupled IVPs that describe the change rate of an important poison, in a nuclear reactor, has been written herein. Specifically, in this article, we have focused on the samarium-149 (as a poison) burnup in a desired pressurized water nuclear reactor and its concentration are given using our MATLAB-linked 'solver'.

  17. IFR starts to burn up weapons-grade material

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    With funding from different parts of the federal government, the Integral Fast Reactor (IFR) project has survived into fiscal year 1994 and is now embarking on a demonstration of how this type of liquid-metal-cooled reactor (LMR) can be used to burn fuel derived from weapons-grade plutonium. This month, an assembly made from weapons-grade material is to be loaded into Experimental Breeder Reactor-II in Idaho, which is serving as the prototype for the IFR concept. Although FY 1994 work is being funded by the DOE, this particular examination of plutonium burnup is backed by the Department of Defense

  18. How Unilever palm oil suppliers are burning up Borneo

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-04-15

    New evidence shows expansion by Unilever palm oil suppliers is driving species extinction in Central Kalimantan, and fuelling climate change. In November 2007, Greenpeace released 'Cooking the Climate', an 82-page report summarizing the findings of a two-year investigation that revealed how the world's largest food, cosmetic and biofuel companies were driving the wholesale destruction of Indonesia's rainforests and peatlands through growing palm oil consumption. This follow-up report provides further evidence of the expansion of the palm oil sector in Indonesia into remaining rainforests, orang-utan habitat and peatlands in Kalimantan. It links the majority of the largest producers in Indonesia to Unilever, probably the largest palm oil corporate consumer in the world.

  19. Innovative methods for the design of miniatured state of the art neutronic channels for future FBR

    International Nuclear Information System (INIS)

    Sivaramakrishna, M.; Shukla, Rahul; Nagaraj, C.P.; Madhusoodanan, K.

    2011-01-01

    Neutron flux in fast reactor varies over 10 decades from shutdown to full power. This is monitored in all states of the reactor by Neutron flux monitoring systems. The system consists of several sets of detectors and instrument channels. For smooth transition from one set of instrument channel to other, interlocks with auto inhibition in safety logic are provided. In the present reactor, Fission counters are placed in control plug under vessel. The signals from these detectors are connected to preamplifiers near to the sensors. The signals from these preamplifiers are again connected to instrument channels in control building local control centres, which generate parameters such as logarithmic power, period, linear power and reactivity signals. Interlocks with auto inhibition are provided in safety logic for smooth transition from one detector-channel to other. With this arrangement, difficulties are faced such as, two sets of detectors are required at two different locations. Three different types of instrument channels are required. In each channel, there is a trade-off between response time and accuracy. Many interlocks with auto inhibitions are to be provided for transition from one system to another. Volume of electronics involved is very high. The pulses are also not periodical. Hence counting techniques do not result in accurate prediction of frequency and intern power. Either the accuracy of counting or the counting time has to be sacrificed. It can also be seen that as the frequency goes up the pulses over lap. Hence estimating the power using pulse counting can't predict the power correctly. In the Campbell method, which is used in the intermediate range also there is a trade-off between the accuracy and response time and linearity could be seen only for 3 decades. Hence the new approach viz. PCA (principle component analysis) is discussed in this paper. PCA is non-parametric method of extracting relevant information from mixed data by reducing a complex

  20. Potentialities of high temperature reactors (HTR)

    International Nuclear Information System (INIS)

    Hittner, D.

    2001-01-01

    This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)