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Sample records for high burn-up cladding

  1. 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)

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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)

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. Development of high performance cladding

    International Nuclear Information System (INIS)

    Kiuchi, Kiyoshi

    2003-01-01

    The developments of superior next-generation light water reactor are requested on the basis of general view points, such as improvement of safety, economics, reduction of radiation waste and effective utilization of plutonium, until 2030 year in which conventional reactor plants should be renovate. Improvements of stainless steel cladding for conventional high burn-up reactor to more than 100 GWd/t, developments of manufacturing technology for reduced moderation-light water reactor (RMWR) of breeding ratio beyond 1.0 and researches of water-materials interaction on super critical pressure-water cooled reactor are carried out in Japan Atomic Energy Research Institute. Stable austenite stainless steel has been selected for fuel element cladding of advanced boiling water reactor (ABWR). The austenite stain less has the superiority for anti-irradiation properties, corrosion resistance and mechanical strength. A hard spectrum of neutron energy up above 0.1 MeV takes place in core of the reduced moderation-light water reactor, as liquid metal-fast breeding reactor (LMFBR). High performance cladding for the RMWR fuel elements is required to get anti-irradiation properties, corrosion resistance and mechanical strength also. Slow strain rate test (SSRT) of SUS 304 and SUS 316 are carried out for studying stress corrosion cracking (SCC). Irradiation tests in LMFBR are intended to obtain irradiation data for damaged quantity of the cladding materials. (M. Suetake)

  11. 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)

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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)

  17. 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.)

  18. 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)

  19. 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.

  20. 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

  1. 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)

  2. 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...

  3. 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.)

  4. 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.

  5. 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)

  6. 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

  7. 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.

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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.)

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. High performance fuel technology development : Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeongyong; Jeong, Y. H.; Park, S. Y.

    2012-04-01

    The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

  18. 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)

  19. 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

  20. 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)

  1. 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)

  2. 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)

  3. Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

    2010-04-01

    The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

  4. 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)

  5. Protective claddings for high strength chromium alloys

    Science.gov (United States)

    Collins, J. F.

    1971-01-01

    The application of a Cr-Y-Hf-Th alloy as a protective cladding for a high strength chromium alloy was investigated for its effectiveness in inhibiting nitrogen embrittlement of a core alloy. Cladding was accomplished by a combination of hot gas pressure bonding and roll cladding techniques. Based on bend DBTT, the cladding alloy was effective in inhibiting nitrogen embrittlement of the chromium core alloy for up to 720 ks (200hours) in air at 1422 K (2100 F). A significant increase in the bend DBTT occurred with longer time exposures at 1422 K or short time exposures at 1589 K (2400 F).

  6. Fundamentals and industrial applications of high power laser beam cladding

    International Nuclear Information System (INIS)

    Bruck, G.J.

    1988-01-01

    Laser beam cladding has been refined such that clad characteristics are precisely determined through routine process control. This paper reviews the state of the art of laser cladding optical equipment, as well as the fundamental process/clad relationships that have been developed for high power processing. Major categories of industrial laser cladding are described with examples chose to highlight particular process attributes

  7. 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.

  8. 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

  9. 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)

  10. 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

  11. 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...

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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

  17. 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

  18. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  19. 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)

  20. 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)

  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. 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)

  3. 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)

  4. 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)

  5. Creep and creep rupture properties of cladding tube (type 316) in high temperature sodium

    International Nuclear Information System (INIS)

    Atsumo, H.

    1977-01-01

    The thin walled small sized seamless AISI 316 steel tubes, which are designated to be domestically used as the fuel cladding tube for sodium cooled fast breeder reactors in Japan, are irradiated in the following sodium of high temperature in the range of 370 deg. C to 700 deg. C, and receive gradually increased internal pressure caused by the fission produced gas generating from the nuclear fuel burn-up inside the cladding tube. Consequently, the creep behavior of fuel cladding tubes under a high temperature sodium environment is an important problem which must be determined and clarified together with their characteristic features under irradiation and in air. In relation to the creep performance of fuel cladding tubes made of AISI 316 steel and other comparable austenitic stainless steels, hardly any studies are found that are made systematically to examine the effect of sodium with sodium purity as parameter or any comparative studies with in-air data at various different temperatures. The present research work was aimed to obtain certain basic design data relating to in-sodium creep performance of the domestic made fuel cladding tubes for fast breeder reactors, and also to gain further date as considered necessary under several sodium conditions. That is, together with establishment of the technology for tensile creep test and internal pressure creep rupture test in flowing sodium of high temperature, a series of tests and studies were performed on the trial made cladding tubes of AISI Type-316 steel. In the first place, two kinds of purity conditions of sodium, close to the actual reactor-operating condition, (oxygen concentration of 10 ppm and 5 ppm respectively) were established, and then uniaxial tensile creep test and rupture test under various temperatures were performed and the resulting data were compared and evaluated against the in-air data. Then, secondly, an internal pressure creep rupture test was conducted under a single purity sodium environment

  6. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    Lott, Randy G.

    2003-01-01

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

  7. Study on the Standard Establishment for the Integrity Assessment of Nuclear Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S-S; Kim, S-H; Jung, Y-K; Yang, C-Y; Kim, I-G; Choi, Y-H; Kim, H-J; Kim, M-W; Rho, B-H [KINS, Daejeon (Korea, Republic of)

    2008-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is the final report.

  8. Study on the standard establishment for the integrity assessment of nuclear fuel cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. S.; Kim, S. H.; Jung, Y. K.; Yang, C. Y.; Kim, I. G.; Choi, Y. H.; Kim, H. J.; Kim, M. W.; Rho, B. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2007-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is 2nd term report.

  9. Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

    International Nuclear Information System (INIS)

    Chatterjee, S.

    2005-01-01

    During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

  10. 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)

  11. 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

  12. 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.

  13. 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

  14. Review and evaluation of cladding attack of LMFBR fuel

    International Nuclear Information System (INIS)

    Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T.

    1977-01-01

    The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio (oxygen to metal ratio), burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature

  15. 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

  16. 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.)

  17. All fiber cladding mode stripper with uniform heat distribution and high cladding light loss manufactured by CO2 laser ablation

    Science.gov (United States)

    Jebali, M. A.; Basso, E. T.

    2018-02-01

    Cladding mode strippers are primarily used at the end of a fiber laser cavity to remove high-power excess cladding light without inducing core loss and beam quality degradation. Conventional manufacturing methods of cladding mode strippers include acid etching, abrasive blasting or laser ablation. Manufacturing of cladding mode strippers using laser ablation consist of removing parts of the cladding by fused silica ablation with a controlled penetration and shape. We present and characterize an optimized cladding mode stripper design that increases the cladding light loss with a minimal device length and manufacturing time. This design reduces the localized heat generation by improving the heat distribution along the device. We demonstrate a cladding mode stripper written on a 400um fiber with cladding light loss of 20dB, with less than 0.02dB loss in the core and minimal heating of the fiber and coating. The manufacturing process of the designed component is fully automated and takes less than 3 minutes with a very high throughput yield.

  18. 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)

  19. 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

  20. 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)

  1. Study on microstructure and high temperature wear resistance of laser cladded nuclear valve clack

    International Nuclear Information System (INIS)

    Zhang Chunliang; Chen Zichen

    2002-01-01

    Laser cladding of Co-base alloy on the nuclear valve-sealing surface are performed with a 5 kW CO 2 transverse flowing laser. The microstructure and the high temperature impact-slide wear resistance of the laser cladded coating and the plasma cladded coating are studied. The results show that the microstructure, the dilution rate and the high temperature impact-slide wear resistance of the laser cladded coating have obvious advantages over the spurt cladding processing

  2. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  3. 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)

  4. 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)

  5. 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)

  6. 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.

  7. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Young Hwan; Park, S. Y.; Lee, M. H.

    2007-04-01

    This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods(LTR) in a commercial reactor

  8. Corrosion characteristics of K-claddings

    International Nuclear Information System (INIS)

    Park, J. Y.; Choi, B. K.; Jung, Y. H.; Jung, Y. H.

    2004-01-01

    The Improvement of the corrosion resistance of nuclear fuel claddings is the critical issue for the successful development of the high burn-up fuel. KAERI have developed the K-claddings having a superior corrosion resistance by controlling the alloying element addition and optimizing the manufacturing process. The comparative evaluation of the corrosion resistance for K-claddings and the foreign claddings was performed and the effect of the heat treatment on the corrosion behavior of K-claddings was also examined. Corrosion tests were carried out in the conditions of 360 .deg. C pure water, PWR-simulating loop and 400 .deg. C steam, From the results of the corrosion tests, it was found that the corrosion resistance of K-claddings is superior to those of Zry4 and A claddings and K6 showed a better corrosion resistance than K3. The corrosion behavior of K-cladding was strongly influenced by the final annealing rather than the intermediate annealing, and the corrosion resistance increased with decreasing the final annealing temperature

  9. Hydrogenation and high temperature oxidation of Zirconium claddings

    International Nuclear Information System (INIS)

    Novotny, T.; Perez-Feró, E.; Horváth, M.

    2015-01-01

    In the last few years a new series of experiments started for supporting the new LOCA criteria, considering the proposals of US NRC. The effects which can cause the embrittlement of VVER fuel claddings were reviewed and evaluated in the framework of the project. The purpose of the work was to determine how the fuel cladding’s hydrogen uptake under normal operating conditions, effect the behavior of the cladding under LOCA conditions. As a first step a gas system equipment with gas valves and pressure gauge was built, in which the zirconium alloy can absorb hydrogen under controlled conditions. In this apparatus E110 (produced by electrolytic method, currently used at Paks NPP) and E110G (produced by a new technology) alloys were hydrogenated to predetermined hydrogen contents. According the results of ring compression tests the E110G alloys lose their ductility above 3200 ppm hydrogen content. This limit can be applied to determine the ductile-brittle transition of the nuclear fuel claddings. After the hydrogenation, high temperature oxidation experiments were carried out on the E110G and E110 samples at 1000 °C and 1200 °C. 16 pieces of E110G and 8 samples of E110 with 300 ppm and 600 ppm hydrogen content were tested. The oxidation of the specimens was performed in steam, under isothermal conditions. Based on the ring compression tests load-displacement curves were recorded. The main objective of the compression tests was to determine the ductile-brittle transition. These results were compared to the results of our previous experiments where the samples did not contain hydrogen. The original claddings showed more ductile behavior than the samples with hydrogen content. The higher hydrogen content resulted in a more brittle mechanical behavior. However no significant difference was observed in the oxidation kinetics of the same cladding types with different hydrogen content. The experiments showed that the normal operating hydrogen uptake of the fuel claddings

  10. 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.

  11. 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

  12. 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

  13. 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.

  14. Fuel clad chemical interactions in fast reactor MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Viswanathan, R., E-mail: rvis@igcar.gov.in

    2014-01-15

    Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel–Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ⋅ [B/(at.% fission)] ⋅ (T/K-705) ⋅ [(O/M)_i-1.935]} + 20.5) for (O/M){sub i} ⩽ 1.98. A new model is proposed for (O/M){sub i} ⩾ 1.98: d/μm = [B/(at.% fission)] ⋅ (T/K-800){sup 0.5} ⋅ [(O/M){sub i}-1.94] ⋅ [P/(W cm{sup −1})]{sup 0.5}. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M){sub i} is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

  15. 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)

  16. 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)

  17. Potential for cladding thermal failure in LWRs during high temperature transients

    International Nuclear Information System (INIS)

    El Genk, M.S.

    1979-01-01

    The temperature increase in the fuel and the cladding during a PCM accident produces film boiling at the cladding surface which may induce zircaloy cladding failure, due to embrittlement, and fuel melting at the centerline of the fuel pellets. Molten fuel may extrude through radial cracks in the fuel and relocate in the fuel-cladding gap. Contact of extruded molten fuel with the cladding, which is at high temperature during film boiling, may induce cladding thermal failure due to melting. An assessment of central fuel melting and molten fuel extrusion into the fuel-cladding gap during a PCM accident is presented. The potential for thermal failure of the zircaloy cladding upon being contacted by molten fuel during such an accident is also analyzed and compared with the applicable experimental evidence

  18. High-resolution characterization of oxidation mechanism of zirconium nuclear fuel cladding alloys

    International Nuclear Information System (INIS)

    Hu, J.; Lozano-Perez, S.; Grovenor, C.

    2015-01-01

    Full text of publication follows. Zirconium alloys are used extensively as cladding materials in modern light water reactors to separate the uranium dioxide (UO 2 ) fuel rods and the coolant water in order to prevent the escape of radioactive fission products whilst maintaining heat transfer to the coolant. With increasing demand for high burn-up in modern nuclear reactors, environmental degradation of these alloys is now the life limiting factor for fuel assemblies. As part of the MUZIC-2 collaboration studying oxidation and hydrogen pickup in Zr alloys, several high resolution analysis techniques have been used to study the microstructure of a range of commercial and developmental Zr alloys. The sample used for this investigation was prepared from a Westinghouse TM developmental alloy with composition of Zr-0.9Nb-0.01Sn-0.08Fe (wt %) in the recrystallized condition. The sample was oxidised in an autoclave at EDF Energy under simulated PWR water conditions at 360 C. degrees for 360 days. Using Transmission Electron Microscope (TEM), we have studied the development of the equiaxed-columnar-equiaxed grain structure, and observe that the columnar grains are both longer and show a stronger preferred texture in more corrosion-resistant alloys. Fresnel imaging revealed the existence of both parallel interconnected pores and some vertically interconnected pores along the columnar oxide grain boundaries, which become more disconnected near the metal-oxide interface. Electron Energy Loss Spectroscopy (EELS) provided accurate quantitative analysis of the oxygen concentration across the interface, identifying the existence of local regions of stoichiometric ZrO and Zr 3 O 2 with varying thickness. These observations will be discussed in the context of current models for oxidation in zirconium alloys. (authors)

  19. 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

  20. 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)

  1. Introduction program of M5TM cladding in Japan

    International Nuclear Information System (INIS)

    Mardon, Jean Paul; Kaneko, Nori

    2008-01-01

    Experience from irradiation in PWR has confirmed that M5 TM possesses all the properties required for upgraded operation including new fuel management approaches and high duty reactor operation. Specifically, the alloy M5 TM has demonstrated impressive improvements over Zircaloy-4 for fuel rod cladding and fuel assembly structural components. Moreover, several irradiation campaigns have been worldwide performed in order to confirm the excellent M5 TM in-pile behavior in very demanding PWR irradiation conditions (high void fraction, heat flux, temperature, lithium content and Zinc injection). Regarding licensing, the authorization for loading M5 TM alloy has been granted by US, UK, South Korean, German, Chinese, South-African, Swedish and Belgian Safety Authorities. Also the French Nuclear Safety Authority has given individually its authorization to load all-M5 TM fuel assembly batches in 1300MWe plants and a generic license to load all-M5 TM fuel in EDF N4 reactors and M5 TM fuel clad in 900MWe reactors for MOX parity fuel management. Licensing is also now underway in Switzerland, Finland, Brazil and Spain. The M5 TM alloy has demonstrated its superiority at burn-ups beyond current licensing limits, through operations in PWR at fuel rod burn-ups exceeding 71GWd/tU in the United States and 78GWd/tU in Europe. The Japanese nuclear industry has planned a stepwise approach to increase the burn-up of the fuel. Step-I fuel (48GWd/tU Fuel Assembly maximum burn-up) which was introduced in the late 80s. In the 90s started the licensing of the Step-II fuel (55GWd/tU Fuel Assembly maximum burn-up). Because the extension of the burn-up is important to reduce discharge fuel and cycle cost, the Japanese industry has plans to further extend the burn-up. In such burn-up region, fuel cladding with even better corrosion properties and very low hydrogen pick-up shall be necessary. M5 TM alloy, with high anticorrosion/hydriding properties, is suitable for not only the Step-II fuel

  2. 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)

  3. 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

  4. 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.)

  5. 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)

  6. In-pile test results of HANA claddings in Halden research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Choi, Byoung Kwon; Jeong, Yong Hwan; Jung, Yun Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    It is a kind of facing tasks in the nuclear industry to develop advanced claddings for high burn-up fuel which is safer and more economical than the existing conventional ones. Since 1997, taking an initiative in KAERI, the Zr cladding development team has carried out the R and D activities for the development of the advanced claddings to be used in the high burn-up fuel (>70,000 MWD.MTU). The team had produced the advanced claddings (HANA, High-performance Alloy for Nuclear Application) from the patented composition and manufacturing process in the international collaboration with U.S. and Japan. Now, the HANA claddings have being demonstrated their good performances from the out-of-pile tests including the corrosion, creep, burst, tensile, microstructures LOCA, RIA, wear, and so on. In parallel to the out-of-pile performance tests, the HANA claddings are being undertaken to evaluate their in-pile properties in Halden research reactor. In this study, it is included the test overviews, conditions, and results of the HANA claddings in the Halden reactor.

  7. Robust cladding light stripper for high-power fiber lasers using soft metals.

    Science.gov (United States)

    Babazadeh, Amin; Nasirabad, Reza Rezaei; Norouzey, Ahmad; Hejaz, Kamran; Poozesh, Reza; Heidariazar, Amir; Golshan, Ali Hamedani; Roohforouz, Ali; Jafari, S Naser Tabatabaei; Lafouti, Majid

    2014-04-20

    In this paper we present a novel method to reliably strip the unwanted cladding light in high-power fiber lasers. Soft metals are utilized to fabricate a high-power cladding light stripper (CLS). The capability of indium (In), aluminum (Al), tin (Sn), and gold (Au) in extracting unwanted cladding light is examined. The experiments show that these metals have the right features for stripping the unwanted light out of the cladding. We also find that the metal-cladding contact area is of great importance because it determines the attenuation and the thermal load on the CLS. These metals are examined in different forms to optimize the contact area to have the highest possible attenuation and avoid localized heating. The results show that sheets of indium are very effective in stripping unwanted cladding light.

  8. Behavior of high burnup fuel rod cladding during long-term dry storage in CASTOR casks

    International Nuclear Information System (INIS)

    Schaberg, A.; Spilker, H.; Goll, W.

    2000-01-01

    Short-time creep and rupture tests were performed to assess the strain potential of cladding of high burnt rods under conditions of dry storage. The tests comprised optimized Zr y-4 cladding samples from fuel rods irradiated to burnups of up to 64 MWd/kg U and were carried out at temperatures of 573 and 643 K at cladding stresses of about 400 and 600 MPa. The stresses, much higher than those occurring in a fuel rod, were chosen to reach circumferential elongations of about 2% within an envisaged testing time of 3-4 days. The creep tests were followed by a low temperature test at 423 K and 100 MPa to assess the long-term behavior of the cladding ductility especially with regard to the effect of a higher hydrogen content in the cladding due to the high burnup. The creep tests showed considerable uniform plastic elongations at these high burnups. It was demonstrated that around 600 K a uniform plastic strain of a least 2% is reached without cladding failure. The low temperature tests at 423 K for up to 5 days revealed no cladding failure under these conditions of reduced cladding ductility. It can be concluded that the increased hydrogen content has no adverse effect on cladding performance. (Authors)

  9. 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)

  10. Measurement and removal of cladding light in high power fiber systems

    Science.gov (United States)

    Walbaum, Till; Liem, Andreas; Schreiber, Thomas; Eberhardt, Ramona; Tünnermann, Andreas

    2018-02-01

    The amount of cladding light is important to ensure longevity of high power fiber components. However, it is usually measured either by adding a cladding light stripper (and thus permanently modifying the fiber) or by using a pinhole to only transmit the core light (ignoring that there may be cladding mode content in the core area). We present a novel noninvasive method to measure the cladding light content in double-clad fibers based on extrapolation from a cladding region of constant average intensity. The method can be extended to general multi-layer radially symmetric fibers, e.g. to evaluate light content in refractive index pedestal structures. To effectively remove cladding light in high power systems, cladding light strippers are used. We show that the stripping efficiency can be significantly improved by bending the fiber in such a device and present respective experimental data. Measurements were performed with respect to the numerical aperture as well, showing the dependency of the CLS efficiency on the NA of the cladding light and implying that efficiency data cannot reliably be given for a certain fiber in general without regard to the properties of the guided light.

  11. 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)

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

    International Nuclear Information System (INIS)

    Siefken, L.J.

    1999-01-01

    Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from above on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown

  17. 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

  18. 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)

  19. 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)

  20. 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

  1. Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee

    International Nuclear Information System (INIS)

    2000-10-01

    The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately

  2. 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

  3. An integrated approach to selecting materials for fuel cladding in advanced high-temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rangacharyulu, C., E-mail: chary.r@usask.ca [Univ. of Saskatchewan, Saskatoon, SK (Canada); Guzonas, D.A.; Pencer, J.; Nava-Dominguez, A.; Leung, L.K.H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    An integrated approach has been developed for selection of fuel cladding materials for advanced high-temperature reactors. Reactor physics, thermalhydraulic and material analyses are being integrated in a systematic study comparing various candidate fuel-cladding alloys. The analyses established the axial and radial neutron fluxes, power distributions, axial and radial temperature distributions, rates of defect formation and helium production using AECL analytical toolsets and experimentally measured corrosion rates to optimize the material composition for fuel cladding. The project has just been initiated at University of Saskatchewan. Some preliminary results of the analyses are presented together with the path forward for the project. (author)

  4. 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)

  5. Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yano, Y., E-mail: yano.yasuhide@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T. [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Ukai, S.; Oono, N. [Materials Science and Engineering, Faculty of Engineering, Hokkaido University, N13, W-8, Kita-ku, Sapporo, Hokkaido, 060-8628 (Japan); Kimura, A. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hayashi, S. [Tokyo Institute of Technology, 2-12-1, Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Torimaru, T. [Nippon Nuclear Fuel Development Co., Ltd., 2163, Narita-cho, Oarai-machi, Ibaraki, 311-1313 (Japan)

    2017-04-15

    Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900–1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

  6. Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

    Science.gov (United States)

    Yano, Y.; Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T.; Ukai, S.; Oono, N.; Kimura, A.; Hayashi, S.; Torimaru, T.

    2017-04-01

    Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900-1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

  7. Steam oxidation of Zr 1% Nb clads of VVER fuels in high temperature

    International Nuclear Information System (INIS)

    Solyanyj, V.I.; Bibilashvili, Yu.K.; Dranenko, V.V.; Levin, A.Ya.; Izrajlevskij, L.B.; Morozov, A.M.

    1984-01-01

    In a wide range of accident conditions processes of clad corrosion effected by steam are rather intensive and in many respects influence the safety of NPP and the after-accident dismantling of a reactor core. This paper discusses the results of comprehensive studies into corrosion behaviour of Zr 1%Nb clads of VVER-type fuels at high temperatures. These studies are a continuation of previous work and the base for the design modelling of corrosion processes

  8. Kilowatt-level cladding light stripper for high-power fiber laser.

    Science.gov (United States)

    Yan, Ping; Sun, Junyi; Huang, Yusheng; Li, Dan; Wang, Xuejiao; Xiao, Qirong; Gong, Mali

    2017-03-01

    We designed and fabricated a high-power cladding light stripper (CLS) by combining a fiber-etched CLS with a cascaded polymer-recoated CLS. The etched fiber reorganizes the numerical aperture (NA) distribution of the cladding light, leading to an increase in the leakage power and a flatter distribution of the leakage proportion in the cascaded polymer-recoated fiber. The index distribution of the cascaded polymer-recoated fiber is carefully designed to ensure an even leakage of cladding light. More stages near the index of 1.451 are included to disperse the heat. The CLS is capable of working consistently under 1187 W of cladding light with an attenuation of 26.59 dB, and the highest local temperature is less than 35°C.

  9. 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)

  10. 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.

  11. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  12. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    International Nuclear Information System (INIS)

    Perez, Emmanuel; Keiser Jr, Dennis D.; Forsmann, Bryan; Janney, Dawn E.; Henley, Jody; Woolstenhulme, Eric C.

    2016-01-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  13. 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)

  14. High Cr ODS steels R and D for high burnup fuel cladding

    International Nuclear Information System (INIS)

    Kimura, A.; Kasada, R.; Kishimoto, H.; Iwata, N.; Cho, H.-S.; Toda, N.; Yutani, K.; Ukai, S.; Fujiwara, M.

    2007-01-01

    High-performance cladding materials is essential to realize highly efficient and high-burnup operation over 150 GWd/t of so called Generation IV nuclear energy systems, such as supercritical-water-cooled reactor (SCWR) and lead-cooled fast reactor (LFR). Oxide dispersion strengthening (ODS) ferritic/ martensitic steels, which contain 9-12%Cr, show rather high resistance to neutron irradiation embrittlement and high strength at elevated temperatures. However, their corrosion resistance is not good enough in SCW and in lead at high temperatures. High-Cr ODS steels have been developed to improve corrosion resistance. An increase in Cr content an addition resulted in a drastic improvement of corrosion resistance in SCW and in lead. On the contrary, high-Cr steels often show an enhancement of aging embrittlement as well as irradiation embrittlement. Anisotropy in tensile properties is another issue. In order to overwhelm these issues, surveillance tests of the material performance have been performed for high Cr-ODS steels produced by new processing technologies. It is demonstrated that the dispersion of nono-sized oxide particles in high density is effective to attain high-performance and high-Cr ODS steels have a high potential as fuel cladding materials for SCWR and LFR with high efficiency and high burnup. (authors)

  15. Amorphous structure evolution of high power diode laser cladded Fe–Co–B–Si–Nb coatings

    International Nuclear Information System (INIS)

    Zhu Yanyan; Li Zhuguo; Huang Jian; Li Min; Li Ruifeng; Wu Yixiong

    2012-01-01

    Highlights: ► Fabricated amorphous composited coating by high power diode laser cladding with single track. ► Lower dilution and higher scanning speed are desired to obtain higher amorphous phase fraction. ► White spots phase with high content of Nb embedded in the amorphous matrix. - Abstract: Fe–Co–B–Si–Nb coatings were fabricated on the surface of low carbon steel using high power diode laser cladding of [(Fe 0.5 Co 0.5 ) 0.75 B 0.2 Si 0.05 ] 95.7 Nb 4.3 amorphous powders at three different scanning speeds of 6, 17 and 50 m/s. At each scanning speed, laser power was optimized to obtain low dilution ratio. Scanning electron microscopy, X-ray diffraction, transmission electron microscopy with energy dispersive spectrometer and electron probe micro analysis were carried out to characterize the microstructure and chemical composition of the cladded coatings. Differential scanning calorimetry was also carried out to investigate the fraction of the amorphous phase. The results showed that dilution ratio and scanning speed were the two main factors for fabricating Fe–Co–B–Si–Nb amorphous coating by high power diode laser cladding. Low dilution ratio was crucial for the formation of amorphous phase. When the dilution ratio was low, the fraction of amorphous phase in the cladded coatings increased upon increasing the scanning speed.

  16. Magnetostrictive clad steel plates for high-performance vibration energy harvesting

    Science.gov (United States)

    Yang, Zhenjun; Nakajima, Kenya; Onodera, Ryuichi; Tayama, Tsuyoki; Chiba, Daiki; Narita, Fumio

    2018-02-01

    Energy harvesting technology is becoming increasingly important with the appearance of the Internet of things. In this study, a magnetostrictive clad steel plate for harvesting vibration energy was proposed. It comprises a cold-rolled FeCo alloy and cold-rolled steel joined together by thermal diffusion bonding. The performances of the magnetostrictive FeCo clad steel plate and conventional FeCo plate cantilevers were compared under bending vibration; the results indicated that the clad steel plate construct exhibits high voltage and power output compared to a single-plate construct. Finite element analysis of the cantilevers under bending provided insights into the magnetic features of a clad steel plate, which is crucial for its high performance. For comparison, the experimental results of a commercial piezoelectric bimorph cantilever were also reported. In addition, the cold-rolled FeCo and Ni alloys were joined by thermal diffusion bonding, which exhibited outstanding energy harvesting performance. The larger the plate volume, the more the energy generated. The results of this study indicated not only a promising application for the magnetostrictive FeCo clad steel plate as an efficient energy harvester, related to small vibrations, but also the notable feasibility for the formation of integrated units to support high-power trains, automobiles, and electric vehicles.

  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. High power operation of cladding pumped holmium-doped silica fibre lasers.

    Science.gov (United States)

    Hemming, Alexander; Bennetts, Shayne; Simakov, Nikita; Davidson, Alan; Haub, John; Carter, Adrian

    2013-02-25

    We report the highest power operation of a resonantly cladding-pumped, holmium-doped silica fibre laser. The cladding pumped all-glass fibre utilises a fluorine doped glass layer to provide low loss cladding guidance of the 1.95 µm pump radiation. The operation of both single mode and large-mode area fibre lasers was demonstrated, with up to 140 W of output power achieved. A slope efficiency of 59% versus launched pump power was demonstrated. The free running emission was measured to be 2.12-2.15 µm demonstrating the potential of this architecture to address the long wavelength operation of silica based fibre lasers with high efficiency.

  2. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  3. Transmission of laser pulses with high output beam quality using step-index fibers having large cladding

    Science.gov (United States)

    Yalin, Azer P; Joshi, Sachin

    2014-06-03

    An apparatus and method for transmission of laser pulses with high output beam quality using large core step-index silica optical fibers having thick cladding, are described. The thick cladding suppresses diffusion of modal power to higher order modes at the core-cladding interface, thereby enabling higher beam quality, M.sup.2, than are observed for large core, thin cladding optical fibers. For a given NA and core size, the thicker the cladding, the better the output beam quality. Mode coupling coefficients, D, has been found to scale approximately as the inverse square of the cladding dimension and the inverse square root of the wavelength. Output from a 2 m long silica optical fiber having a 100 .mu.m core and a 660 .mu.m cladding was found to be close to single mode, with an M.sup.2=1.6. Another thick cladding fiber (400 .mu.m core and 720 .mu.m clad) was used to transmit 1064 nm pulses of nanosecond duration with high beam quality to form gas sparks at the focused output (focused intensity of >100 GW/cm.sup.2), wherein the energy in the core was laser pulses was about 6 ns. Extending the pulse duration provided the ability to increase the delivered pulse energy (>20 mJ delivered for 50 ns pulses) without damaging the silica fiber.

  4. 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

  5. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi

    2001-03-01

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  6. 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)

  7. High-performace cladding-pumped erbium-doped fibre laser and amplifier

    International Nuclear Information System (INIS)

    Kotov, L V; Likhachev, M E; Bubnov, M M; Medvedkov, O I; Lipatov, D S; Vechkanov, N N; Guryanov, Aleksei N

    2012-01-01

    We report cladding-pumped erbium-doped fibre laser and amplifier configurations. Through fibre design optimisation, we have achieved a record-high laser slope efficiency, 40 % with respect to absorbed pump power (λ = 976 nm), and an output power of 7.5 W. The erbium-doped fibre amplifier efficiency reaches 32 %.

  8. 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)

  9. Cladding glass ceramic for use in high powered lasers

    Science.gov (United States)

    Marker, Alexander J.; Campbell, John H.

    1998-01-01

    A Cu-doped/Fe-doped low expansion glass ceramic composition comprising in Wt. %: SiO{sub 2} 50--65; Al{sub 2}O{sub 3} 18--27; P{sub 2}O{sub 5} 0--10; Li{sub 2}O 2--6; Na{sub 2}O 0--2; K{sub 2}O 0--2; B{sub 2}O{sub 3} 0--1; MgO 0--4; ZnO 0--5; CaO 0--4; BaO 0--5; TiO{sub 2} 1--3; ZrO{sub 3} 1--3; As{sub 2}O{sub 3} 0--1.5; Sb{sub 2}O{sub 3} 0--1.5; CuO 0--3; and Fe{sub 2}O{sub 3} 0--1 wherein the total amount of SiO{sub 2}, Al{sub 2}O{sub 3} and P{sub 2}O{sub 5} is 80--89 wt. %, and said glass ceramic contains as a dopant 0.1--3 wt. % CuO, 0.1--1 wt. % Fe{sub 2}O{sub 3} or a combined CuO+Fe{sub 2}O{sub 3} amount of 0.1--4 wt. %. The glass ceramic composition is suitable for use as a cladding material for solid laser energy storage mediums as well as for use in beam attenuators for measuring laser energy level and beam blocks or beam dumps used for absorbing excess or unused laser energy.

  10. 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)

  11. 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.

  12. 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)

  13. Yb-doped phosphate double-cladding optical fiber for high-power laser applications

    Science.gov (United States)

    Mura, E.; Scarpignato, G. C.; Lousteau, J.; Boetti, N. G.; Abrate, S.; Milanese, D.

    2013-02-01

    A Yb-doped phosphate glass double cladding optical fiber was prepared using a custom designed glass composition (P2O5 - Al2O3 - Li2O - B2O3 - BaO - PbO - La2O3) for high-power amplifier and laser applications. The preform drawing method was followed, with the preform being fabricated using the rotational casting technique. This technique, previously developed for tellurite, fluoride or chalcogenide glass preforms is reported for the first time using rare earth doped phosphate glasses. The main challenge was to design an adequate numerical aperture between first and second cladding while maintaining similar thermo-mechanical properties in view of the fiber drawing process. The preform used for the fiber drawing was produced by rod-in-tube technique at a rotation speed of 3000 rpm. The rotational casting technique allowed the manufacturing of an optical fiber featuring high quality interfaces between core and internal cladding and between the internal and external cladding, respectively. Loss attenuation was measured using the cut-back method and lasing was demonstrated at 1022 nm by core pumping with a fiber pigtailed laser diode at the wavelength of 976 nm.

  14. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    International Nuclear Information System (INIS)

    Chung, H.M.

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 μm in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307 degree C rather than the normal 288 degree C, a relatively thick (50 to 70 μm) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs

  15. 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)

  16. Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

    2009-01-01

    The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

  17. Volume Bragg grating narrowed high-power and highly efficient cladding-pumped Raman fiber laser.

    Science.gov (United States)

    Liu, Jun; Yao, Weichao; Zhao, Chujun; Shen, Deyuan; Fan, Dianyuan

    2014-12-10

    High-power and highly efficient operation of a single-mode cladding-pumped Raman fiber laser with narrow lasing bandwidth is demonstrated. The spectral narrowing was realized by an external cavity containing a volume Bragg grating with a center wavelength of 1658 nm. A maximum output power of 10.4 W at 1658.3 nm with a spectral linewidth (FWHM) of ∼0.1  nm was obtained for the launched pump power of 18.4 W, corresponding to a slope efficiency of 109% with respect to the launched pump power. Lasing characteristics of free-running operation are also evaluated and discussed.

  18. High Temperature Dry Sliding Friction and Wear Performance of Laser Cladding WC/Ni Composite Coating

    Directory of Open Access Journals (Sweden)

    YANG Jiao-xi

    2016-06-01

    Full Text Available Two different types of agglomerate and angular WC/Ni matrix composite coatings were deposited by laser cladding. The high temperature wear resistance of these composite coatings was tested with a ring-on-disc MMG-10 apparatus. The morphologies of the worn surfaces were observed using a scanning electron microscopy (SEM equipped with an energy dispersive spectroscopy (EDS for elemental composition. The results show that the high temperature wear resistance of the laser clad WC/Ni-based composite coatings is improved significantly with WC mass fraction increasing. The 60% agglomerate WC/Ni composite coating has optimal high temperature wear resistance. High temperature wear mechanism of 60% WC/Ni composite coating is from abrasive wear of low temperature into composite function of the oxidation wear and abrasive wear.

  19. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    Chosson, Raphael

    2014-01-01

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

  20. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    International Nuclear Information System (INIS)

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  1. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    Languille, A.

    1979-07-01

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  2. 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)

  3. 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

  4. 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)

  5. 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)

  6. Study on the improvement of nuclear fuel cladding reliability

    International Nuclear Information System (INIS)

    Rheem, Karp Soon; Han, Jung Ho; Jeong, Yong Hwan; Lee, Deok Hyun

    1987-12-01

    In order to improve the nuclear fuel cladding reliability for high burn-up fuels, the corrosion resistance of laser beam surface treated and β-quenched zircaloys and the mechanical characteristics including fatigue, burst, and out-of-pile PCMI characteristics of heat treated zircaloys were investigated. In addition, the inadiation characteristics of Ko-Ri reactor fuel claddings was examined. It was found that the wasteside corrosion resistance of commercial zircaloys was improved remarkably by laser beam surface treatment. The out-of-pile transient cladding failures were investigated in terms of hoop stress versus time-to-failures by means of mandrel loading units at 25 deg C and 325 deg C. Fatigue characteristics of the β-quenched and as-received zircaloy cladding were investigated by using an internal oil pressurization method which can simulate the load-following operation cycle. The results were in good agreement with the existing data obtained by conventional methods for commercial zircaloys. Burst tests were performed with commercial and the β-quenched zircaloys in high pressure argon gas atmosphere as a function of burst temperature. The burst stress decreased linearly in the α phase region up to 600 deg C and hereafter the decrement of the burst stress decreased gradually with temperature in the β-phase region. For the first time, the burst characteristic of the irradiated zircaloy-4 cladding tubes released from Ko-Ri nuclear power unit 1 was investigated, and attempts were made to trace the cause of cladding failures by examining the failed structure and fret marks by debris. (Author)

  7. 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.

  8. 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

  9. 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.

  10. Parametric Study and Multi-Criteria Optimization in Laser Cladding by a High Power Direct Diode Laser

    Science.gov (United States)

    Farahmand, Parisa; Kovacevic, Radovan

    2014-12-01

    In laser cladding, the performance of the deposited layers subjected to severe working conditions (e.g., wear and high temperature conditions) depends on the mechanical properties, the metallurgical bond to the substrate, and the percentage of dilution. The clad geometry and mechanical characteristics of the deposited layer are influenced greatly by the type of laser used as a heat source and process parameters used. Nowadays, the quality of fabricated coating by laser cladding and the efficiency of this process has improved thanks to the development of high-power diode lasers, with power up to 10 kW. In this study, the laser cladding by a high power direct diode laser (HPDDL) as a new heat source in laser cladding was investigated in detail. The high alloy tool steel material (AISI H13) as feedstock was deposited on mild steel (ASTM A36) by a HPDDL up to 8kW laser and with new design lateral feeding nozzle. The influences of the main process parameters (laser power, powder flow rate, and scanning speed) on the clad-bead geometry (specifically layer height and depth of the heat affected zone), and clad microhardness were studied. Multiple regression analysis was used to develop the analytical models for desired output properties according to input process parameters. The Analysis of Variance was applied to check the accuracy of the developed models. The response surface methodology (RSM) and desirability function were used for multi-criteria optimization of the cladding process. In order to investigate the effect of process parameters on the molten pool evolution, in-situ monitoring was utilized. Finally, the validation results for optimized process conditions show the predicted results were in a good agreement with measured values. The multi-criteria optimization makes it possible to acquire an efficient process for a combination of clad geometrical and mechanical characteristics control.

  11. Development of Diffusion barrier coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Allen, Todd; Cole, James

    2013-02-27

    The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.

  12. Proposal of 99.99%-aluminum/7N01-Aluminum clad beam tube for high energy booster of Superconducting Super Collider

    International Nuclear Information System (INIS)

    Ishimaru, Hajime

    1994-01-01

    Proposal of 99.99% pure aluminum/7N01 aluminum alloy clad beam tube for high energy booster in Superconducting Super Collider is described. This aluminum clad beam tube has many good performances, but a eddy current effect is large in superconducting magnet quench collapse. The quench test result for aluminum clad beam tube is basically no problem against magnet quench collapse. (author)

  13. 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

  14. Laser cladding process development for high carbon steel substrates

    CSIR Research Space (South Africa)

    Lengopeng, T

    2014-11-01

    Full Text Available 316L as a butter layer did not lead to the formation of hard brittle chromium carbides even though the chromium content is high enough for the formation of these stable carbides. It was however observed that the use of an AISI 316L butter layer did...

  15. Widely tunable asymmetric long-period fiber grating with high sensitivity using optical polymer on laser-ablated cladding.

    Science.gov (United States)

    Chen, Nan-Kuang; Hsu, Der-Yi; Chi, Sien

    2007-08-01

    We demonstrate high-efficiency, wideband-tunable, laser-ablated long-period fiber gratings that use an optical polymer overlay. Portions of the fiber cladding are periodically removed by CO(2) laser pulses to induce periodic index changes for coupling the core mode into cladding modes. An optical polymer with a high thermo-optic coefficient with a dispersion distinct from that of silica is used on a deep-ablated cladding structure so that the effective indices of cladding modes become dispersive and the resonant wavelengths can be efficiently tuned. The tuning efficiency can be as high as 15.8 nm/ degrees C, and the tuning range can be wider than 105 nm (1545-1650 nm).

  16. Synthesis and Characterization of High-Entropy Alloy AlFeCoNiCuCr by Laser Cladding

    Directory of Open Access Journals (Sweden)

    Xiaoyang Ye

    2011-01-01

    Full Text Available High-entropy alloys have been recently found to have novel microstructures and unique properties. In this study, a novel AlFeCoNiCuCr high-entropy alloy was prepared by laser cladding. The microstructure, chemical composition, and constituent phases of the synthesized alloy were characterized by SEM, EDS, XRD, and TEM, respectively. High-temperature hardness was also evaluated. Experimental results demonstrate that the AlFeCoNiCuCr clad layer is composed of only BCC and FCC phases. The clad layers exhibit higher hardness at higher Al atomic content. The AlFeCoNiCuCr clad layer exhibits increased hardness at temperature between 400–700°C.

  17. 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)

  18. 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)

  19. Geometry modeling of single track cladding deposited by high power diode laser with rectangular beam spot

    Science.gov (United States)

    Liu, Huaming; Qin, Xunpeng; Huang, Song; Hu, Zeqi; Ni, Mao

    2018-01-01

    This paper presents an investigation on the relationship between the process parameters and geometrical characteristics of the sectional profile for the single track cladding (STC) deposited by High Power Diode Laser (HPDL) with rectangle beam spot (RBS). To obtain the geometry parameters, namely cladding width Wc and height Hc of the sectional profile, a full factorial design (FFD) of experiment was used to conduct the experiments with a total of 27. The pre-placed powder technique has been employed during laser cladding. The influence of the process parameters including laser power, powder thickness and scanning speed on the Wc and Hc was analyzed in detail. A nonlinear fitting model was used to fit the relationship between the process parameters and geometry parameters. And a circular arc was adopted to describe the geometry profile of the cross-section of STC. The above models were confirmed by all the experiments. The results indicated that the geometrical characteristics of the sectional profile of STC can be described as the circular arc, and the other geometry parameters of the sectional profile can be calculated only using Wc and Hc. Meanwhile, the Wc and Hc can be predicted through the process parameters.

  20. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure in the pr......Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... to analyze, compare, and discuss how these various construction solutions point out strategies for development based on fundamentally different mindsets. The research questions address the following issues: How to learn from traditional construction principles: When do we see limitations of tectonic maneuver......, to ask for more restrictive building codes. As an example, in Denmark there are series of increasing demands in the current building legislations that are focused at enhancing the energy performance of buildings, which consequently foster rigid insulation standards and ask for improvement of air...

  1. High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments

    International Nuclear Information System (INIS)

    Chung, H.M.; Garde, A.M.; Kassner, T.F.

    1977-01-01

    The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350 0 C

  2. Assessment of effect of Yb3+ ion pairs on a highly Yb-doped double-clad fibre laser

    Science.gov (United States)

    Vallés, J. A.; Martín, J. C.; Berdejo, V.; Cases, R.; Álvarez, J. M.; Rebolledo, M. Á.

    2018-03-01

    Using a previously validated characterization method based on the careful measurement of the characteristic parameters and fluorescence emission spectra of a highly Yb-doped double-clad fibre, we evaluate the contribution of ion pair induced processes to the output power of a double-clad Yb-doped fibre ring laser. This contribution is proved to be insignificant, contrary to analysis by other authors, who overestimate the role of ion pairs.

  3. Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

    International Nuclear Information System (INIS)

    Zmitko, M.

    1999-01-01

    In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

  4. High-Temperature Tolerance in Multi-Scale Cermet Solar-Selective Absorbing Coatings Prepared by Laser Cladding.

    Science.gov (United States)

    Pang, Xuming; Wei, Qian; Zhou, Jianxin; Ma, Huiyang

    2018-06-19

    In order to achieve cermet-based solar absorber coatings with long-term thermal stability at high temperatures, a novel single-layer, multi-scale TiC-Ni/Mo cermet coating was first prepared using laser cladding technology in atmosphere. The results show that the optical properties of the cermet coatings using laser cladding were much better than the preplaced coating. In addition, the thermal stability of the optical properties for the laser cladding coating were excellent after annealing at 650 °C for 200 h. The solar absorptance and thermal emittance of multi-scale cermet coating were 85% and 4.7% at 650 °C. The results show that multi-scale cermet materials are more suitable for solar-selective absorbing coating. In addition, laser cladding is a new technology that can be used for the preparation of spectrally-selective coatings.

  5. High-Temperature Tolerance in Multi-Scale Cermet Solar-Selective Absorbing Coatings Prepared by Laser Cladding

    Directory of Open Access Journals (Sweden)

    Xuming Pang

    2018-06-01

    Full Text Available In order to achieve cermet-based solar absorber coatings with long-term thermal stability at high temperatures, a novel single-layer, multi-scale TiC-Ni/Mo cermet coating was first prepared using laser cladding technology in atmosphere. The results show that the optical properties of the cermet coatings using laser cladding were much better than the preplaced coating. In addition, the thermal stability of the optical properties for the laser cladding coating were excellent after annealing at 650 °C for 200 h. The solar absorptance and thermal emittance of multi-scale cermet coating were 85% and 4.7% at 650 °C. The results show that multi-scale cermet materials are more suitable for solar-selective absorbing coating. In addition, laser cladding is a new technology that can be used for the preparation of spectrally-selective coatings.

  6. 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)

  7. 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)

  8. 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%.

  9. 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.

  10. 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)

  11. Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

  12. 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)

  13. High Hydrogen Content Graphene Hydride Compounds & High Cross-Section Cladding Coatings for Fast Neutron Detection

    International Nuclear Information System (INIS)

    Chandrashekhar, MVS

    2017-01-01

    The objective is to develop and implement a superior low-cost, large area (potentially >32in), easily deployable, close proximity, harsh environment innovative neutron sensor needed for next generation fuel cycle monitoring. We will exploit recent breakthroughs at the PI's lab on the electrochemistry of epitaxial graphene (EG) formed on commercial SiC wafers, a transformative nanomaterial system with superior radiation detection and durability properties to develop a new paradigm in detection for fast neutrons, a by-product of fission reactors. There are currently few effective detection/monitoring schemes, especially solid-state ones at present. This is essential for monitoring and control of future fuel cycles to make them more efficient and reliable. By exploiting these novel materials, as well as innovative hybrid SiC/EG/Cladding device architectures conceived by the team, will develop low-cost, high performance solutions to fast-neutron detection. Finally, we will also explore 3-terminal device implementations for neutron detectors with built-in electronic gain to further shrink these devices and improve their sensitivity.

  14. 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.)

  15. Microstructure and Wear Behavior of CoCrFeMnNbNi High-Entropy Alloy Coating by TIG Cladding

    Directory of Open Access Journals (Sweden)

    Wen-yi Huo

    2015-01-01

    Full Text Available Alloy cladding coatings are widely prepared on the surface of tools and machines. High-entropy alloys are potential replacements of nickel-, iron-, and cobalt-base alloys in machining due to their excellent strength and toughness. In this work, CoCrFeMnNbNi HEA coating was produced on AISI 304 steel by tungsten inert gas cladding. The microstructure and wear behavior of the cladding coating were studied by X-ray diffraction, scanning electron microscopy, energy dispersive spectrometer, microhardness tester, pin-on-ring wear tester, and 3D confocal laser scanning microscope. The microstructure showed up as a nanoscale lamellar structure matrix which is a face-centered-cubic solid solution and niobium-rich Laves phase. The microhardness of the cladding coating is greater than the structure. The cladding coating has excellent wear resistance under the condition of dry sliding wear, and the microploughing in the worn cladding coating is shallower and finer than the worn structure, which is related to composition changes caused by forming the nanoscale lamellar structure of Laves phase.

  16. Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1976-04-01

    The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800 0 F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests

  17. Study of the solidification of M2 high speed steel Laser Cladding coatings

    Directory of Open Access Journals (Sweden)

    Candel, J. J.

    2013-10-01

    Full Text Available High speed steel laser cladding coatings are complex because cracks appear and the hardness is lower than expected. In this paper AISI M2 tool steel coatings on medium carbon AISI 1045 steel substrate have been manufactured and after Laser Cladding (LC processing it has been applied a tempering heat treatment to reduce the amount of retained austenite and to precipitate secondary carbides. The study of metallurgical transformations by Scanning Electron Microscopy (SEM and Electron Back Scattered Diffraction (EBSD shows that the microstructure is extremely fine and complex, with eutectic transformations and MC, M2C and M6C precipitation. Therefore, after the laser coating is necessary to use post-weld heat treatments.Los recubrimientos de acero rápido por Laser Cladding (LC son complejos porque aparecen fisuras y la dureza es menor a la esperada. En este trabajo se han fabricado recubrimientos de acero AISI M2 sobre acero al carbono AISI 1045 y tras el procesado por láser, se han revenido para reducir la cantidad de austenita retenida y precipitar carburos secundarios. El estudio de las transformaciones metalúrgicas con Microscopía Electrónica de Barrido (MEB y Difracción de Electrones Retrodispersados (EBSD muestra que la microestructura es extremadamente fina y compleja, presenta transformaciones eutécticas y precipitación de carburos MC, M2C y M6C. Por tanto, tras el recubrimiento por láser es necesario recurrir a tratamientos térmicos post-soldeo.

  18. 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.)

  19. 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.

  20. 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

  1. Performance of high burned PWR fuel during transient

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio

    1992-01-01

    In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

  2. CO2 laser-fabricated cladding light strippers for high-power fiber lasers and amplifiers.

    Science.gov (United States)

    Boyd, Keiron; Simakov, Nikita; Hemming, Alexander; Daniel, Jae; Swain, Robert; Mies, Eric; Rees, Simon; Andrew Clarkson, W; Haub, John

    2016-04-10

    We present and characterize a simple CO2 laser processing technique for the fabrication of compact all-glass optical fiber cladding light strippers. We investigate the cladding light loss as a function of radiation angle of incidence and demonstrate devices in a 400 μm diameter fiber with cladding losses of greater than 20 dB for a 7 cm device length. The core losses are also measured giving a loss of cladding light stripping of a 300 W laser diode with minimal heating of the fiber coating and packaging adhesives.

  3. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  4. 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

  5. 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.

  6. 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)

  7. 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)

  8. 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)

  9. 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)

  10. Repository emplacement costs for Al-clad high enriched uranium spent fuel

    International Nuclear Information System (INIS)

    McDonell, W.R.; Parks, P.B.

    1994-01-01

    A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs

  11. Electrophoretic Deposition for the Fabrication of High-Performance Metal-Ceramic Hybrid Cladding

    International Nuclear Information System (INIS)

    Park, Junghwan; Jung, Yangil; Park, Dongjun; Kim, Hyungil; Park, Jeongyong; Koo, Yanghyun

    2014-01-01

    Metal-ceramic hybrid cladding consisting of a Zr liner and SiC f /SiC composite is one of the candidate systems. To achieve a high-performance metal-ceramic hybrid cladding, it is important to synthesize the SiC f /SiC composites with high flexural strength. The most common interphases, such as pyrolytic carbon (PyC) and boron nitride (BN) coating, have been applied on the surface of SiC fibers by chemical vapor deposition (CVD) or chemical vapor infiltration (CVI). In addition, the SiC matrix phase for SiC f /SiC composites has been commonly formed by CVI and polymer infiltration and pyrolysis (PIP), which are very costly and complicated processes. For this reason, the fabrication process of SiC f /SiC composites that is low-cost and simple has been strongly needed. In this study, weak phase coating using a commercial colloidal carbon black suspension was performed on SiC fibers through electrophoretic deposition (EPD), and carbon-coated SiC f /SiC composites were fabricated by EPD. The mechanical properties at room temperature were evaluated to investigate the effect of the carbon interfacial layer on the mechanical properties of carbon-coated SiC f /SiC composites. In this study, it was concluded that the EPD method is effective for homogeneous carbon black coating on SiC fibers, and that the carbon coating layer on SiC fibers plays an important role in optimizing the interface between fibers and the matrix, and enhances the toughness of carbon-coated SiC f /SiC composites during fracture

  12. Thermal and irradiation effects on high-temperature mechanical properties of materials for SCWR fuel cladding

    International Nuclear Information System (INIS)

    Kano, F.; Tsuchiya, Y.; Oka, K.

    2009-01-01

    The thermal and irradiation effects on high-temperature mechanical properties are examined for candidate alloys for fuel cladding of supercritical water-cooled reactors (SCRWs). JMTR (Japan Materials Testing Reactor) and Experimental Fast Reactor JOYO were utilized for neutron irradiation tests, considering their fluence and temperature. Irradiation was performed with JMTR at 600degC up to 4x10 24 n/m 2 and with JOYO at 600degC and 700degC up to 6x10 25 n/m 2 . Tensile test, creep test and hardness measurement were carried out for high-temperature mechanical properties. Based on the uniaxial creep test, the extrapolation curves were drawn with time-temperature relationships utilizing the Larson and Miller Parameter. Several candidate alloys are expected to satisfy the design requirement from the estimation of the creep rupture stress for 50000 hours. Comparing the creep strengths under irradiated and unirradiated conditions, it was inferred that creep deformation was dominated by the thermal effect rather than the irradiation at SCWR core condition. The microstructure was examined using transmission electron microscope (TEM) analysis, focusing on void swelling and helium (He) bubble formation. Void formation was observed in the materials irradiated with JOYO at 600degC but not at 700degC. However, its effect on the deformation of components was estimated to be tolerable since their size and density were negligibly small. The manufacturability of the thin-wall, small-diameter tube was confirmed for the potential candidate alloys through the trial tests in the factory where the fuel cladding tube is manufactured. (author)

  13. Electrophoretic Deposition for the Fabrication of High-Performance Metal-Ceramic Hybrid Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Junghwan; Jung, Yangil; Park, Dongjun; Kim, Hyungil; Park, Jeongyong; Koo, Yanghyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Metal-ceramic hybrid cladding consisting of a Zr liner and SiC{sub f}/SiC composite is one of the candidate systems. To achieve a high-performance metal-ceramic hybrid cladding, it is important to synthesize the SiC{sub f}/SiC composites with high flexural strength. The most common interphases, such as pyrolytic carbon (PyC) and boron nitride (BN) coating, have been applied on the surface of SiC fibers by chemical vapor deposition (CVD) or chemical vapor infiltration (CVI). In addition, the SiC matrix phase for SiC{sub f}/SiC composites has been commonly formed by CVI and polymer infiltration and pyrolysis (PIP), which are very costly and complicated processes. For this reason, the fabrication process of SiC{sub f}/SiC composites that is low-cost and simple has been strongly needed. In this study, weak phase coating using a commercial colloidal carbon black suspension was performed on SiC fibers through electrophoretic deposition (EPD), and carbon-coated SiC{sub f}/SiC composites were fabricated by EPD. The mechanical properties at room temperature were evaluated to investigate the effect of the carbon interfacial layer on the mechanical properties of carbon-coated SiC{sub f}/SiC composites. In this study, it was concluded that the EPD method is effective for homogeneous carbon black coating on SiC fibers, and that the carbon coating layer on SiC fibers plays an important role in optimizing the interface between fibers and the matrix, and enhances the toughness of carbon-coated SiC{sub f}/SiC composites during fracture.

  14. 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

  15. 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)

  16. 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

  17. 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)

  18. High-temperature oxidation kinetics of sponge-based E110 cladding alloy

    Science.gov (United States)

    Yan, Yong; Garrison, Benton E.; Howell, Mike; Bell, Gary L.

    2018-02-01

    Two-sided oxidation experiments were recently conducted at 900°C-1200 °C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates in a relatively short time, the new sponge-based E110 demonstrated steam oxidation behavior comparable to Zircaloy-4. Sample weight gain and oxide layer thickness measurements were performed on oxidized E110 specimens and compared to oxygen pickup and oxide layer thickness calculations using the Cathcart-Pawel correlation. Our study shows that the sponge-based E110 follows the parabolic law at temperatures above 1015 °C. At or below 1015 °C, the oxidation rate was very low when compared to Zircaloy-4 and can be represented by a cubic expression. No breakaway oxidation was observed at 1000 °C for oxidation times up to 10,000 s. Arrhenius expressions are given to describe the parabolic rate constants at temperatures above 1015 °C and cubic rate constants are provided for temperatures below 1015 °C. The weight gains calculated by our equations are in excellent agreement with the measured sample weight gains at all test temperatures. In addition to the as-fabricated E110 cladding sample, prehydrided E110 cladding with hydrogen concentrations in the 100-150 wppm range was also investigated. The effect of hydrogen content on sponge-based E110 oxidation kinetics was minimal. No significant difference was found between as-fabricated and hydrided samples with regard to oxygen pickup and oxide layer thickness for hydrogen contents below 150 wppm.

  19. Prediction of failure of highly irradiated Zircaloy clad tubes under reactivity initiated accidents

    International Nuclear Information System (INIS)

    Jernkvist, L.O.

    2003-01-01

    This paper deals with failure of irradiated Zircaloy tubes under the heat-up stage of a reactivity initiated accident (RIA). More precisely, by use of a model for plastic strain localization and necking failure, we theoretically analyse the effects of local surface defects on clad ductility and survivability under RIA. The results show that even very shallow surface defects, e.g. arising from a non-uniform or partially spilled oxide layer, have a strong limiting effect on clad ductility. Moreover, in presence of surface defects, the ability of the clad tube to expand radially without necking failure is found to be extremely sensitive to the stress biaxiality ratio σ zz /σ θθ , which is here assumed to be in the range from 0 to 1. The results of our analysis are compared with clad ductility data available in literature, and their consequences for clad failure prediction under RIA are discussed. In particular, the results raise serious concerns regarding the applicability of failure criteria, which are based on clad strain energy density. These criteria do not capture the observed sensitivity to stress biaxiality on clad failure propensity. (author)

  20. Cladding creepdown under compression

    International Nuclear Information System (INIS)

    Hobson, D.O.

    1977-01-01

    Light-water power reactors use Zircaloy tubing as cladding to contain the UO 2 fuel pellets. In-service operating conditions impose an external hydrostatic force on the cladding, causing it to creep down into eventual contact with the fuel. Knowledge of the rate of such creepdown is of great importance to modelers of fuel element performance. An experimental system was devised for studying creepdown that meets several severe requirements by providing (1) correct stress state, (2) multiple positions for measuring radial displacement of the cladding surface, (3) high-precision data, and (4) an experimental configuration compact enough to fit in-reactor. A microcomputer-controlled, eddy-current monitoring system was developed for this study and has proven highly successful in measuring cladding deformation with time at temperatures of 371 0 C (700 0 F) and higher, and at pressures as high as 21 MPa

  1. High temperature oxidation test of oxide dispersion strengthened (ODS) steel claddings

    International Nuclear Information System (INIS)

    Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasushi

    2006-07-01

    In a feasibility study of ODS steel cladding, its high temperature oxidation resistance was evaluated. Although addition of Cr is effective for preventing high temperature oxidation, excessively higher amount of Cr leads to embrittlement due to the Cr-rich α' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, high temperature oxidation test was conducted for ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) 9Cr-ODS martensitic and 12Cr-ODS ferritic steel have superior high temperature oxidation resistance compared to 11mass%Cr PNC-FMS and even 17mass% SUS430 and equivalent to austenitic PNC316. (2) The superior oxidation resistance of ODS steel was attributed to earlier formation of the protective alpha-Cr 2 O 3 layer at the matrix and inner oxide scale interface. The grain size of ODS steel is finer than that of PNC-FMS, so the superior oxidation resistance of ODS steel can be attributed to the enhanced Cr-supplying rate throughout the accelerated grain boundary diffusion. Finely dispersed Y 2 O 3 oxide particles in the ODS steel matrix may also stabilized the adherence between the protective alpha-Cr 2 O 3 layer and the matrix. (author)

  2. BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

    International Nuclear Information System (INIS)

    Pohl, P.; Wehle, F.

    2010-01-01

    Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

  3. Experimental and numerical investigation on cladding of corrosion-erosion resistant materials by a high power direct diode laser

    Science.gov (United States)

    Farahmand, Parisa

    In oil and gas industry, soil particles, crude oil, natural gas, particle-laden liquids, and seawater can carry various highly aggressive elements, which accelerate the material degradation of component surfaces by combination of slurry erosion, corrosion, and wear mechanisms. This material degradation results into the loss of mechanical properties such as strength, ductility, and impact strength; leading to detachment, delamination, cracking, and ultimately premature failure of components. Since the failure of high valued equipment needs considerable cost and time to be repaired or replaced, minimizing the tribological failure of equipment under aggressive environment has been gaining increased interest. It is widely recognized that effective management of degradation mechanisms will contribute towards the optimization of maintenance, monitoring, and inspection costs. The hardfacing techniques have been widely used to enhance the resistance of surfaces against degradation mechanisms. Applying a surface coating improves wear and corrosion resistance and ensures reliability and long-term performance of coated parts. A protective layer or barrier on the components avoids the direct mechanical and chemical contacts of tool surfaces with process media and will reduce the material loss and ultimately its failure. Laser cladding as an advanced hardfacing technique has been widely used for industrial applications in order to develop a protective coating with desired material properties. During the laser cladding, coating material is fused into the base material by means of a laser beam in order to rebuild a damaged part's surface or to enhance its surface function. In the hardfacing techniques such as atmospheric plasma spraying (APS), high velocity oxygen-fuel (HVOF), and laser cladding, mixing of coating materials with underneath surface has to be minimized in order to utilize the properties of the coating material most effectively. In this regard, laser cladding offers

  4. Numerical Ballooning and Burst Prediction of Fuel Cladding During LOCA Transients in LWR

    International Nuclear Information System (INIS)

    Landau, E.; Weiss, Y.; Szanto, M.

    2014-01-01

    Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis, most former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. Several codes for reactor core and fuel rod analysis exist nowadays, such as FRAPTRAN1.4 or RELAP5-3D . These are tailor-made codes, designed to predict general core behavior and fuel performance, and while they are also used in predicting core components behavior during accident conditions, including those of cladding ballooning and failure with good accuracy, they contain several limitations on modeling the full transient cladding thermo mechanical phenomena. Limitations such as mechanical models being one dimensional or in axisymmetric geometries only, relying mostly on analytical models therefore having further restricting assumptions in return for accuracy, etc. These limitations disable the simulation of several important aspects, such as modeling 3D azimuthal behavior for example. The objective of the current work is to develop a comprehensive numerical model for predicting zircalloy cladding thermo mechanical behavior during a LOCA. The model will eventually predicts full cladding ballooning and burst behavior followed by fuel relocation, for fuel rods that can be subjected to 3D distributed flux. The model is fully three dimensional and is created using the commercial FEM numerical simulation software ABAQUS© applying

  5. High-temperature steam oxidation kinetics of the E110G cladding alloy

    International Nuclear Information System (INIS)

    Király, Márton; Kulacsy, Katalin; Hózer, Zoltán; Perez-Feró, Erzsébet; Novotny, Tamás

    2016-01-01

    In the course of recent years, several experiments were performed at MTA EK (Centre for Energy Research, Hungarian Academy of Sciences) on the isothermal high-temperature oxidation of the improved Russian cladding alloy E110G in steam/argon atmosphere. Using these data and designing additional supporting experiments, the oxidation kinetics of the E110G alloy was investigated in a wide temperature range, between 600 °C and 1200 °C. For short durations (below 500 s) or high temperatures (above 1065 °C) the oxidation kinetics was found to follow a square-root-of-time dependence, while for longer durations and in the intermediate temperature range (800–1000 °C) it was found to approach a cube-root-of-time dependence rather than a square-root one. Based on the results a new best-estimate and a conservative oxidation kinetics model were created. - Highlights: • Steam oxidation kinetics of E110G was studied at MTA EK based on old and new data. • New best-estimate and conservative steam oxidation kinetics were proposed for E110G. • The exponent of oxidation time changed depending on oxidation temperature. • A simple exponential curve was used instead of Arrhenius-type curve for the factor.

  6. 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

  7. 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.)

  8. 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

  9. 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

  10. 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

  11. Progress In Developing an Impermeable, High Temperature Ceramic Composite for Advanced Reactor Clad And Structural Applications

    International Nuclear Information System (INIS)

    Feinroth, Herbert; Hao, Bernard; Fehrenbacher, Larry; Patterson, Mark

    2002-01-01

    Most Advanced Reactors for Energy and Space Applications require higher temperature materials for fuel cladding and core internal structures. For temperatures above 500 deg. C, metal alloys do not retain sufficient strength or long term corrosion resistance for use in either water, liquid metal or gas cooled systems. In the case of water cooled systems, such metals react exo-thermically with water during core overheating accidents, thus requiring extensive and expensive emergency systems to protect against major releases. Past efforts to apply ceramic composites (oxide, carbide or nitride based) having passive safety characteristics, good strength properties at high temperatures, and reasonable resistance to crack growth, have not been successful, either because of irradiation induced effects, or lack of impermeability to fission gases. Under a Phase 1 SBIR (Small Business Innovative Research) project sponsored by DOE's Office of Nuclear Energy, the authors have developed a new material system that may solve these problems. A hybrid tubular structure (0.6 inches in outside diameter) consisting of an inner layer of monolithic silicon carbide (SiC) and outer layers of SiC-SiC composite, bonded to the inner layer, has been fabricated in small lengths. Room temperature permeability tests demonstrate zero gas leakage at pressures up to 120 psig internal pressure. Four point flexural bending tests on these hybrid tubular specimens demonstrate a 'graceful' failure mode: i.e. - the outer composite structure sustains a failure mode under stress that is similar to the yield vs. stress characteristics of metal structures. (authors)

  12. High slope efficiency and high refractive index change in direct-written Yb-doped waveguide lasers with depressed claddings.

    Science.gov (United States)

    Palmer, Guido; Gross, Simon; Fuerbach, Alexander; Lancaster, David G; Withford, Michael J

    2013-07-15

    We report the first Yb:ZBLAN and Yb:IOG10 waveguide lasers fabricated by the fs-laser direct-writing technique. Pulses from a Titanium-Sapphire laser oscillator with 5.1 MHz repetition rate were utilized to generate negative refractive index modifications in both glasses. Multiple modifications were aligned in a depressed cladding geometry to create a waveguide. For Yb:ZBLAN we demonstrate high laser slope efficiency of 84% with a maximum output power of 170 mW. By using Yb:IOG10 a laser performance of 25% slope efficiency and 72 mW output power was achieved and we measured a remarkably high refractive index change exceeding Δn = 2.3 × 10(-2).

  13. Corrosion performance of optimised and advanced fuel rod cladding in PWRs at high burnups

    International Nuclear Information System (INIS)

    Jourdain, P.; Hallstadius, L.; Pati, S.R.; Smith, G.P.; Garde, A.M.

    1997-01-01

    The corrosion behaviour both in-pile and out-of-pile for a number of cladding alloys developed by ABB to meet the current and future needs for fuel rod cladding with improved corrosion resistance is presented. The cladding materials include: 1) Zircaloy-4 (OPTIN) with optimised composition and processing and Zircaloy-2 optimised for Pressurised Water Reactors (PWR), (Zircaloy-2P), and 2) several alternative zirconium-based alloys with compositions outside the composition range for Zircaloys. The data presented originate from fuel rods irradiated in six PWRs to burnups up to about 66 MWd/kgU and from tests conducted in 360 o water autoclave. Also included are in-pile fuel rod growth measurements on some of the alloys. (UK)

  14. Highly scalable, resonantly cladding-pumped, Er-doped fiber laser with record efficiency.

    Science.gov (United States)

    Dubinskii, M; Zhang, J; Ter-Mikirtychev, V

    2009-05-15

    We report the performance of a resonantly cladding-pumped, Yb-free, Er-doped fiber laser. We believe this is the first reported resonantly cladding-pumped fiber-Bragg-grating-based, Er-doped, large-mode-area (LMA) fiber laser. The laser, pumped by fiber-coupled InGaAsP/InP laser diode modules at 1,532.5 nm, delivers approximately 48 W of cw output at 1,590 nm. It is believed to be the highest power ever reported from a Yb-free Er-doped LMA fiber. This fully integrated laser also has the optical-to-optical efficiency of approximately 57%, to the best of our knowledge, the highest efficiency reported for cladding-pumped unidirectionally emitting Er-doped laser.

  15. Modeling of mechanical behavior of quenched zirconium-based nuclear fuel claddings after a high temperature oxidation

    International Nuclear Information System (INIS)

    Cabrera-Salcedo, A.

    2012-01-01

    During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirconium-based fuel claddings undergo a high temperature oxidation (up to 1200 C), then a water quench. After a single-side steam oxidation followed by a direct quench, the cladding is composed of three layers: an oxide (Zirconia) outer layer (formed at HT), always brittle at Room Temperature (RT), an intermediate oxygen stabilized alpha layer, always brittle at RT, called alpha(O), and an inner 'prior-beta' layer, which is the only layer able to keep some significant Post Quench (PQ) ductility at RT. However, hydrogen absorbed because of service exposure or during the LOCA transient, concentrates in this layer and may leads to its embrittlement. To estimate the PQ mechanical properties of these materials, Ring Compression Tests (RCT) are widely used because of their simplicity. Small sample size makes RCTs advantageous when a comparison with irradiated samples is required. Despite their good reproducibility, these tests are difficult to interpret as they often present two or more load drops on the engineering load-displacement curve. Laboratories disagree about their interpretation. This study proposes an original fracture scenario for a stratified PQ cladding tested by RCT, and its associated FE model. Strong oxygen content gradient effect on layers mechanical properties is taken into account in the model. PQ thermal stresses resulting from water quench of HT oxidized cladding are investigated, as well as progressive damage of three layers during an RCT. The proposed scenario is based on interrupted RCT analysis, post- RCT sample's outer layers observation for damage evaluation, RCTs of prior-beta single-layer rings, and mechanical behavior of especially chemically adjusted samples. The force displacement curves appearance is correctly reproduced using the obtained FE model. The proposed fracture scenario elucidates RCTs of quenched zirconium-based nuclear fuel

  16. Improvement of fuel-element reliability by insertion of UO2 microspheres in the gap between pellet and clad

    International Nuclear Information System (INIS)

    Mehedinteanu, S.; Glodeanu, F.; Dobos, I.

    1979-01-01

    With the accumulation of power reactor fuel operating experience, the study of the PCI phenomenon and the development of remedies have become important items in fuel research and development everywhere. The 'power-ramp' failure has drawn attention to the problem of obtaining high reliability from high burn-up fuel rods. Considerable attention has been paid to minimizing the cladding stresses imparted by fuel pellets during the power ramp. The paper describes a new concept of pellet-clad bonding by insertion of UO 2 microspheres in the gap. It is pointed out that the main advantages of this concept are: the low friction coefficient between pellet and clad; the accomodation of cracked pellet expansion by local microyielding of irradiation-embrittled clad; the reduced ridge height by use of undished pellets or other pellet shape; that the fine-sized UO 2 microspheres infiltrate around the pellets thus permitting the use of cracked or chipped pellets and also sintered pellets without the previously required grinding step needed for accurate sizing, etc. (author)

  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. Effects of pellet-to-cladding gap design parameters on the reliability of high burnup PWR fuel rods under steady state and transient conditions

    International Nuclear Information System (INIS)

    Tas, Fatma Burcu; Ergun, Sule

    2013-01-01

    Highlights: • Fuel performance of a typical Pressurized Water Reactor rod is analyzed. • Steady state fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • Transient fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • The optimum pellet to cladding gap thickness and gap gas pressure values of the simulated fuel are determined. • The effects of pellet to cladding gap design parameters on nuclear fuel reliability are examined. - Abstract: As an important improvement in the light water nuclear reactor operations, the nuclear fuel burnup rate is increased in recent decades and this increase causes heavier duty for the nuclear fuel. Since the high burnup fuel is exposed to very high thermal and mechanical stresses and since it operates in an environment with high radiation for about 18 month cycles, it carries the risk of losing its integrity. In this study; it is aimed to determine the effects of pellet–cladding gap thickness and gap pressure on reliability of high burnup nuclear fuel in Pressurized Water Reactors (PWRs) under steady state operation conditions and suggest optimum values for the examined parameters only and validate these suggestions for a transient condition. In the presented study, fuel performance was analyzed by examining the effects of pellet–cladding gap thickness and gap pressure on the integrity of high burnup fuels. This work is carried out for a typical Westinghouse type PWR fuel. The steady state conditions were modeled and simulated with FRAPCON-3.4a steady state fuel performance code and the FRAPTRAN-1.4 fuel transient code was used to calculate transient fuel behavior. The analysis included the changes in the important nuclear fuel design limitations such as the centerline temperature, cladding stress, strain and oxidation with the change in pellet–cladding gap thickness and initial pellet–cladding gap gas

  19. 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.)

  20. Laser Cladding of CPM Tool Steels on Hardened H13 Hot-Work Steel for Low-Cost High-Performance Automotive Tooling

    Science.gov (United States)

    Chen, J.; Xue, L.

    2012-06-01

    This paper summarizes our research on laser cladding of high-vanadium CPM® tool steels (3V, 9V, and 15V) onto the surfaces of low-cost hardened H13 hot-work tool steel to substantially enhance resistance against abrasive wear. The results provide great potential for fabricating high-performance automotive tooling (including molds and dies) at affordable cost. The microstructure and hardness development of the laser-clad tool steels so obtained are presented as well.

  1. A new high temperature deformation model for zircaloy clad ballooning under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Brzoska, B.; Cheliotis, G.; Kunick, A.; Senski, G.

    1977-01-01

    Assuming Zircaloy clad ballooning occurs predominantly by thermal activated secondary creep, generally a power law is applied to describe the creep rate analytically. According to Norton the creep rate is taken as a power function of the cladding hoop stress multiplied by a numerical constant which is determined by the cladding structural properties and a Boltzmann factor including the creep activation energy, the gas constant and the cladding temperature respectively. As is well known, the stress exponent is not a constant value in the total range of LOCA stresses, but increases steadily with stress. This difficulty is avoided by introducing into the Norton law a plastic flow-factor including a limiting stress, which was derived by G. Senski using plastic crack models from Dugdale and Irwin. For LOCA applications the limiting stress is identified with the burst stress, which is experimentally determined. A total number of about 280 directly heated KWU burst tests including two types of experiments: (i) controlled temperature transient tests, (ii) creep rupture tests, are used to fit the burst stress of KWU zircaloy tubes simulating the whole range of LOCA temperatur

  2. Thermal creep behavior of N36 zirconium alloy cladding tube

    International Nuclear Information System (INIS)

    Wang, P.; Zhao, W.; Dai, X.

    2015-01-01

    N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

  3. Development of oxide dispersion strengthened 9Cr ferritic-martensitic steel clad tube for fast reactor

    International Nuclear Information System (INIS)

    Laha, K.; Saroja, S.; Mathew, M.D.; Jayakumar, T.; Vijay, R.; Venugopal Reddy, A.; Lakshminarayana, B.; Kapoor, Komal; Jha, S.K.; Tonpe, S.S.

    2012-01-01

    One of the key issues in the economical operation of FBR is to achieve high burn-up of fuel (200-250 GWd/t) which considerably reduces the fuel cycle cost. This imposes stringent requirements of void swelling resistance upto 200 dpa for the core structural materials. Presently used alloy 09 (a modified austenitic stainless steel, 15Cr-15Ni-Ti) for PFBR has void swelling limit less than 150 dpa. Because of the inherent void swelling resistance, 9-12Cr steels ferritic/martensitic steels are qualified for irradiation upto 200 dpa but their low creep strength at temperatures above 600 deg C restricts their application as a clad material. Oxide dispersion strengthening is found to be promising means of extending the creep resistance of ferritic/martensitic steels beyond 650 deg C without sacrificing the inherent advantages of high thermal conductivity and low swelling of ferritic steels

  4. 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.

  5. 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)

  6. 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

  7. 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)

  8. 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)

  9. Diffusion in cladding materials

    International Nuclear Information System (INIS)

    Anand, M.S.; Pande, B.M.; Agarwala, R.P.

    1992-01-01

    Aluminium has been used as a cladding material in most research reactors because its low neutron absorption cross section and ease of fabrication. However, it is not suitable for cladding in power reactors and as such zircaloy-2 is normally used as a clad because it can withstand high temperature. It has low neutron absorption cross section, good oxidation, corrosion, creep properties and possesses good mechanical strength. With the passage of time, further development in this branch of science took place and designers started looking for better neutron economy and less hydrogen pickup in PHW reactors. The motion of fission products in the cladding material could pose a problem after long operation. In order to understand their behaviour under reactor environment, it is essential to study first the diffusion under normal conditions. These studies will throw light on the interaction of defects with impurities which would in turn help in understanding the mechanism of diffusion. In this article, it is intended to discuss the diffusion behaviour of impurities in cladding materials.(i.e. aluminium, zircaloy-2, zirconium-niobium alloy etc.). (author). 94 refs., 4 figs., 3 tabs

  10. Cladding mode coupling in highly localized fiber Bragg gratings: modal properties and transmission spectra.

    Science.gov (United States)

    Thomas, Jens; Jovanovic, Nemanja; Becker, Ria G; Marshall, Graham D; Withford, Michael J; Tünnermann, Andreas; Nolte, Stefan; Steel, M J

    2011-01-03

    The spectral characteristics of a fiber Bragg grating (FBG) with a transversely inhomogeneous refractive index profile, differs considerably from that of a transversely uniform one. Transmission spectra of inhomogeneous and asymmetric FBGs that have been inscribed with focused ultrashort pulses with the so-called point-by-point technique are investigated. The cladding mode resonances of such FBGs can span a full octave in the spectrum and are very pronounced (deeper than 20dB). Using a coupled-mode approach, we compute the strength of resonant coupling and find that coupling into cladding modes of higher azimuthal order is very sensitive to the position of the modification in the core. Exploiting these properties allows precise control of such reflections and may lead to many new sensing applications.

  11. 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)

  12. 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)

  13. 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.)

  14. 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)

  15. 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)

  16. Air oxidation of Zircaloy-4, M5 (registered) and ZIRLOTM cladding alloys at high temperatures

    International Nuclear Information System (INIS)

    Steinbrueck, M.; Boettcher, M.

    2011-01-01

    The paper presents the results of isothermal and transient oxidation experiments of the advanced cladding alloys M5 (registered) and ZIRLO TM in comparison to Zircaloy-4 in air at temperatures from 973 to 1853 K. Generally, oxidation in air leads to a strong degradation of the cladding material. The main mechanism of this process is the formation of zirconium nitride and its re-oxidation. From the point of view of safety, the barrier effect of the fuel cladding is lost much earlier than during accident transients with a steam atmosphere only. Comparison of the three alloys investigated reveals a qualitatively similar, but quantitatively varying oxidation behavior in air. The mainly parabolic oxidation kinetics, where applicable, is comparable for the three alloys. Strong differences of up to 500% in oxidation rates were observed after transition to linear kinetics at temperatures below 1300 K. The paper presents kinetic rate constants as well as critical times and oxide scale thicknesses at the point of transition from parabolic to linear kinetics.

  17. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  18. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

    2012-11-05

    This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

  19. Microstructural Characterization of Clad Interface in Welds of Ni-Cr-Mo High Strength Low Alloy Steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong-Eun; Kim, Min-Chul; Lee, Ho-Jin; Kim, Keong-Ho [KAERI, Daejeon (Korea, Republic of); Lee, Ki-Hyoung [KAIST, Daejeon (Korea, Republic of); Lee, Chang-Hee [Hanyang Univ., Seoul (Korea, Republic of)

    2011-08-15

    SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at 610°C for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

  20. Microstructural Characterization of Clad Interface in Welds of Ni-Cr-Mo High Strength Low Alloy Steel

    International Nuclear Information System (INIS)

    Kim, Hong-Eun; Kim, Min-Chul; Lee, Ho-Jin; Kim, Keong-Ho; Lee, Ki-Hyoung; Lee, Chang-Hee

    2011-01-01

    SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at 610°C for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

  1. High Hydrogen Content Graphene Hydride Compounds & High Cross-­ Section Cladding Coatings for Fast Neutron Detection

    Energy Technology Data Exchange (ETDEWEB)

    Chandrashekhar, MVS [Univ. of South Carolina, Columbia, SC (United States)

    2017-06-21

    The objective is to develop and implement a superior low-cost, large area (potentially >32in), easily deployable, close proximity, harsh environment innovative neutron sensor needed for next generation fuel cycle monitoring. We will exploit recent breakthroughs at the PI’s lab on the electrochemistry of epitaxial graphene (EG) formed on commercial SiC wafers, a transformative nanomaterial system with superior radiation detection and durability properties to develop a new paradigm in detection for fast neutrons, a by-product of fission reactors. There are currently few effective detection/monitoring schemes, especially solid-state ones at present. This is essential for monitoring and control of future fuel cycles to make them more efficient and reliable. By exploiting these novel materials, as well as innovative hybrid SiC/EG/Cladding device architectures conceived by the team, will develop low-cost, high performance solutions to fast-neutron detection. Finally, we will also explore 3-terminal device implementations for neutron detectors with built-in electronic gain to further shrink these devices and improve their sensitivity.

  2. 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.

  3. 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)

  4. 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)

  5. 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)

  6. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  7. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H.

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  8. Critical cladding radius for hybrid cladding modes

    Science.gov (United States)

    Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

    2018-05-01

    In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

  9. Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

    International Nuclear Information System (INIS)

    Mazeres, Benoit

    2013-01-01

    One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

  10. High temperature behaviour of E110G and E110 fuel claddings in various mixtures of steam and air

    International Nuclear Information System (INIS)

    Perez-Feró, Erzsébet; Novotny, Tamás; Horváth, Márta; Kunstár, Mihály; Vér, Nóra; Hózer, Zoltán

    2014-01-01

    Experiments with sponge base E110G and the traditional E110 were carried out to compare the oxidation kinetics of these alloys in steam, in hydrogen rich steam, in steam-air and in air atmosphere and to study the effect of hydrogen- and nitrogen-containing environment on the oxidation. The effect of oxidizing atmosphere on the mechanical behaviour of the claddings was also investigated. The new and the traditional types of cladding rings were oxidised at high temperature (600°C – 1200°C). Oxidation of both alloys in steam-air mixture and in air atmosphere resulted in faster oxidation kinetics compared to steam. In many cases bumpy, porous oxide layer have been found. The presence of hydrogen in the steam atmosphere had no significant effect on the oxidation kinetics. Comparing the two alloys, more favourable behaviour of oxidised E110G was observed regarding the oxidation kinetics, breakaway oxidation and load bearing capability in all cases. (author)

  11. Effect of laser power on clad metal in laser-TIG combined metal cladding

    Science.gov (United States)

    Utsumi, Akihiro; Hino, Takanori; Matsuda, Jun; Tasoda, Takashi; Yoneda, Masafumi; Katsumura, Munehide; Yano, Tetsuo; Araki, Takao

    2003-03-01

    TIG arc welding has been used to date as a method for clad welding of white metal as bearing material. We propose a new clad welding process that combines a CO2 laser and a TIG arc, as a method for cladding at high speed. We hypothesized that this method would permit appropriate control of the melted quantity of base metal by varying the laser power. We carried out cladding while varying the laser power, and investigated the structure near the boundary between the clad layer and the base metal. Using the laser-TIG combined cladding, we found we were able to control appropriately the degree of dilution with the base metal. By applying this result to subsequent cladding, we were able to obtain a clad layer of high quality, which was slightly diluted with the base metal.

  12. LOCA testing of high burnup PWR fuel in the HBWR. Additional PIE on the cladding of the segment 650-5

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Espeland, M.; Jenssen, H.K.

    2008-07-01

    IFA-650.5, a test with pre-irradiated fuel in the Halden Project LOCA test series, was conducted on October 23rd, 2006. The fuel rod had been used in a commercial PWR and had a high burnup, 83 MWd/kgU. Experimental arrangements of the fifth test were similar to the preceding LOCA tests. The peak cladding temperature (PCT) level was higher than in the third and fourth tests, 1050 C. A peak temperature close to the target was achieved and cladding burst occurred at approx. 750 C. Within the joint programme framework of the Halden Project PIE was done, consisting of gamma scanning, visual inspection, neutron-radiography, hydrogen analysis and metallography / ceramography. An additional extensive PIE including metallography, hydrogen analysis, and hardness measurements of cross-sections at seven axial elevations was done. It was completed to study the high burnup and LOCA induced effects on the Zr-4 cladding, namely the migration of oxygen into the cladding from the inside surface, the cladding distension, and the burst (author)(tk)

  13. 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

  14. 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)

  15. Thermally Stable Siloxane Hybrid Matrix with Low Dielectric Loss for Copper-Clad Laminates for High-Frequency Applications.

    Science.gov (United States)

    Kim, Yong Ho; Lim, Young-Woo; Kim, Yun Hyeok; Bae, Byeong-Soo

    2016-04-06

    We report vinyl-phenyl siloxane hybrid material (VPH) that can be used as a matrix for copper-clad laminates (CCLs) for high-frequency applications. The CCLs, with a VPH matrix fabricated via radical polymerization of resin blend consisting of sol-gel-derived linear vinyl oligosiloxane and bulky siloxane monomer, phenyltris(trimethylsiloxy)silane, achieve low dielectric constant (Dk) and dissipation factor (Df). The CCLs with the VPH matrix exhibit excellent dielectric performance (Dk = 2.75, Df = 0.0015 at 1 GHz) with stability in wide frequency range (1 MHz to 10 GHz) and at high temperature (up to 275 °C). Also, the VPH shows good flame resistance without any additives. These results suggest the potential of the VPH for use in high-speed IC boards.

  16. Conditioning of high activity solid waste: fuel claddings and dissolution residues

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    This chapter reports on experimental studies of embedding into matrix material, the melting and conversion of zircaloy, and waste properties and characterization. Methods are developed for embedding the waste scrap into a solid and resistant matrix material in order to confine the radioactivity and to prevent it from dispersion. The matrix materials investigated are lead alloys, ceramics and compacted graphite or aluminium powder. The treatment of fuel hulls by melting or chemical conversion is described. Cemented hulls are characterized and the pyrophoricity of zircaloy fines is investigated. Topics considered include the embedding of hulls into graphite and aluminium, the embedding of hulls and dissolution residues into alumino-ceramics, the solidification of alpha-bearing wastes into a ceramic matrix, and the conditioning of cladding waste by eutectoidic melting and by embedding in glass

  17. High-power, cladding-pumped all-fiber laser with selective transverse mode generation property.

    Science.gov (United States)

    Li, Lei; Wang, Meng; Liu, Tong; Leng, Jinyong; Zhou, Pu; Chen, Jinbao

    2017-06-10

    We demonstrate, to the best of our knowledge, the first cladding-pumped all-fiber oscillator configuration with selective transverse mode generation based on a mode-selective fiber Bragg grating pair. Operating in the second-order (LP 11 ) mode, maximum output power of 4.2 W is obtained with slope efficiency of about 38%. This is the highest reported output power of single higher-order transverse mode generation in an all-fiber configuration. The intensity distribution profile and spectral evolution have also been investigated in this paper. Our work suggests the potential of realizing higher power with selective transverse mode operation based on a mode-selective fiber Bragg grating pair.

  18. High stability of polymer optical fiber with dye doped cladding for illumination systems

    Energy Technology Data Exchange (ETDEWEB)

    Jaramillo-Ochoa, L. [Centro de Física Aplicada y Tecnología Avanzada, Universidad Nacional Autónoma de México, Boulevard Juriquilla 3001, Querétaro 76230, México (Mexico); Narro-García, R. [Centro de Física Aplicada y Tecnología Avanzada, Universidad Nacional Autónoma de México, Boulevard Juriquilla 3001, Querétaro 76230, México (Mexico); Universidad Autónoma de Chihuahua, Facultad de Ingeniería, Circuito Universitario S/N, 31125 Chihuahua, Chih., México (Mexico); Ocampo, M.A. [Centro de Física Aplicada y Tecnología Avanzada, Universidad Nacional Autónoma de México, Boulevard Juriquilla 3001, Querétaro 76230, México (Mexico); Quintero-Torres, R., E-mail: rquintero@fata.unam.mx [Centro de Física Aplicada y Tecnología Avanzada, Universidad Nacional Autónoma de México, Boulevard Juriquilla 3001, Querétaro 76230, México (Mexico)

    2017-04-15

    In this work, the photodegradation of a polymer optical fiber with Rhodamine doped cladding as a function of illumination time and excitation intensity is presented. To show the effect of photodegradation on different bulk geometries and environments, the photodegradation from a dye doped preform and a PMMA thick film is also evaluated. The reversible and the irreversible degradation of the florescent material were quantified under an established excitation scheme. To this purpose, a four-level system to model the photodegradation rates and its relation with the population of the states is presented and it is used to justify a possible underlying mechanism. The obtained results suggest an increase of one order of magnitude in the stability (lifetime) of the polymer optical fiber with respect to the preform or the thick film geometry stability.

  19. Investigation of the high temperature steam oxidation of Zircaloy 4 cladding tubes

    International Nuclear Information System (INIS)

    Leistikow, S.; Berg, H. v.; Kraft, R.; Pott, E.; Schanz, G.

    1979-01-01

    Also for the ORNL Zircaloy 4 cladding material, an intermediate decrease of the proportion of the ZrO 2 /α-phase layer was found, followed by an drastic increase when the breakaway of the ZrO 2 -scale occurred. Other reasons for small divergencies were evaluated, for instance temperature and time measurements, metallographic evaluation of layer thicknesses, consequences of one-sided (ORNL) and double-sided (KfK) oxidation. The so-called anomalous effect of steam oxidation during temperature transients was reproduced qualitatively and-in case that a reduced gain of oxygen was observed-explained by the predominant existence of the monoclinic oxide phase. The creep-rupture tests below 800 0 C showed a moderate prolongation of time-to-rupture when the tests were performed in steam (or after preoxidation in steam) instead of argon. Also slightly reduced maximum circumferential strain could be measured. (orig./RW) [de

  20. 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

  1. Microstructure and high temperature oxidation resistance of Ti-Ni gradient coating on TA2 titanium alloy fabricated by laser cladding

    Science.gov (United States)

    Liu, Fencheng; Mao, Yuqing; Lin, Xin; Zhou, Baosheng; Qian, Tao

    2016-09-01

    To improve the high temperature oxidation resistance of TA2 titanium alloy, a gradient Ni-Ti coating was laser cladded on the surface of the TA2 titanium alloy substrate, and the microstructure and oxidation behavior of the laser cladded coating were investigated experimentally. The gradient coating with a thickness of about 420-490 μm contains two different layers, e.g. a bright layer with coarse equiaxed grain and a dark layer with fine and columnar dendrites, and a transition layer with a thickness of about 10 μm exists between the substrate and the cladded coating. NiTi, NiTi2 and Ni3Ti intermetallic compounds are the main constructive phases of the laser cladded coating. The appearance of these phases enhances the microhardness, and the dense structure of the coating improves its oxidation resistance. The solidification procedure of the gradient coating is analyzed and different kinds of solidification processes occur due to the heat dissipation during the laser cladding process.

  2. 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

  3. 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.

  4. 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

  5. Evaluation of a Ductility after High Temperature Oxidation with the Three-Point Bend Test in Zirconium Alloys

    International Nuclear Information System (INIS)

    Jung, Yang Il; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan

    2010-01-01

    In a light water reactor, the fuel cladding play an important role of preventing leakage of radioactive materials into the coolant, and thus the mechanical integrity of the cladding should be guaranteed under the conditions of normal and transient operation. In the case of a loss of coolant accident (LOCA), the cladding is subjected to a high temperature oxidation which is finally quenched because of an emergency coolant reflooding into the core. In this situation, the current LOCA criteria consist of five separate requirements: i) peak cladding temperature, ii) maximum cladding oxidation, iii) maximum hydrogen generation, iv) coolable geometry, and v) long-term cooling. The claddings lose their ductility due to the microstructural phase transformation from beta to martensite alpha-prime. and hydrogen up-take after LOCA. Since the reduction in ductility can induce embrittlement of claddings, post-quench ductility is one of the major concerns in transient operation circumstances. For the analysis, usually ring compression test are performed on ring samples cut from the tube to examine the oxidized cladding ductility. However, the test would not be applicable to the platelet samples which are general form of a specimen for developing alloys. As a high burn-up fuel cladding materials, Zircaloys are being replaced by modern zirconium alloys such as ZIRLO, and M5. Korea has also developed a new fuel cladding material HANA (high performance alloy for nuclear application) by the Korea Atomic Energy Research Institute. Because of the different composition of the newer claddings in comparison with the conventional Zircaloy-4, the high temperature oxidation behavior and the ductility after the oxidation would be different, and the properties should be evaluated how much the newer claddings were improved

  6. 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)

  7. 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

  8. Microstructure and wear resistance of laser cladded Ni-Cr-Co-Ti-V high-entropy alloy coating after laser remelting processing

    Science.gov (United States)

    Cai, Zhaobing; Cui, Xiufang; Liu, Zhe; Li, Yang; Dong, Meiling; Jin, Guo

    2018-02-01

    An attempt, combined with the technologies of laser cladding and laser remelting, has been made to develop a Ni-Cr-Co-Ti-V high entropy alloy coating. The phase composition, microstructure, micro-hardness and wear resistance (rolling friction) were studied in detail. The results show that after laser remelting, the phase composition remains unchanged, that is, as-cladded coating and as-remelted coatings are all composed of (Ni, Co)Ti2 intermetallic compound, Ti-rich phase and BCC solid solution phase. However, after laser remelting, the volume fraction of Ti-rich phase increases significantly. Moreover, the micro-hardness is increased, up to ∼900 HV at the laser remelting parameters: laser power of 1 kW, laser spot diameter of 3 mm, and laser speed of 10 mm/s. Compared to the as-cladded high-entropy alloy coating, the as-remelted high-entropy alloy coatings have high friction coefficient and low wear mass loss, indicating that the wear resistance of as-remelted coatings is improved and suggesting practical applications, like coatings on brake pads for wear protection. The worn surface morphologies show that the worn mechanism of as-cladded and as-remelted high-entropy alloy coatings are adhesive wear.

  9. High-temperature air oxidation of E110 and Zr-1%Nb alloys claddings with coatings

    International Nuclear Information System (INIS)

    Kuprin, A.S.; Belous, V.A.; Voyevodin, V.N.; Bryk, V.V.; Vasilenko, R.L.; Ovcharenko, V.D.; Tolmachova, G.N.; V'yugov, P.N.

    2014-01-01

    Results of experimental study of the influence of protective vacuum-arc claddings on the base of compounds zirconium-chromium and of its nitrides on air oxidation resistance at temperatures 660, 770, 900, 1020, 1100 deg C during 3600 s. of tubes produced of zirconium alloys E110 and Zr-1%Nb (calcium-thermal alloy of Ukrainian production) are presented. Change of hardness, the width of oxide layer and depth of oxygen penetration into alloys from the side of coating and without coating are investigated by the methods of nanoindentation and by scanning electron microscopy. It is shown that the thickness of oxide layer in zirconium alloys at temperatures 1020 and 1100 deg C from the side of the coating doesn't exceed 5 μm, and from the unprotected side reaches the value of ≥ 120 μm with porous and rough structure. Tubes with coatings save their shape completely independently of the type of alloy; tubes without coatings deform with the production of through cracks

  10. Research of the mode instability threshold in high power double cladding Yb-doped fiber amplifiers

    International Nuclear Information System (INIS)

    Wang, Yanshan; Ma, Yi; Sun, Yinhong; Peng, Wanjing; Tang, Chun; Liu, Qinyong; Ke, Weiwei; Wang, Xiaojun

    2017-01-01

    We experimentally investigate the behavior of the mode instability (MI) threshold in the double cladding Yb-doped fiber amplifier when the amplifier is pumped by broad linewidth laser diodes and narrow linewidth laser diodes respectively. It is found that the MI threshold increases by 26% when the amplifier is pumped by the broad linewidth laser diodes. Experiment results show that the MI threshold is affected by the local heat load rather than the average or the total heat load. The calculation shows that the local heat deposit actually plays the key role to stimulate the MI behaviour. At the MI threshold position in the fiber, the local heat deposit also changes dramatically. The effect of the thermal conductivity on the MI threshold is also studied. Our investigation shows that the MI threshold increases from 1269 W to 1950 W when the thermal conductivity of the fiber amplifier is increased from 0.3 W/(m . K) to 5 W/(m . K). Through optimizing the pump linewidth and the cooling efficiency of the gain fiber, the MI threshold is doubled in our experiment. (copyright 2017 by WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  11. Research of the mode instability threshold in high power double cladding Yb-doped fiber amplifiers

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yanshan; Ma, Yi; Sun, Yinhong; Peng, Wanjing; Tang, Chun [Institute of Applied Electronics, CAEP, Mianyang, Sichuan (China); The Key Laboratory of Science and Technology on High Energy Laser, CAEP, Mianyang, Sichuan (China); Liu, Qinyong; Ke, Weiwei [Institute of Applied Physics and Computational Mathematics, CAEP, Beijing (China); Wang, Xiaojun [Institute of Applied Physics and Computational Mathematics, CAEP, Beijing (China); Technical Institute of Physics and Chemistry, CAS, Beijing (China)

    2017-08-15

    We experimentally investigate the behavior of the mode instability (MI) threshold in the double cladding Yb-doped fiber amplifier when the amplifier is pumped by broad linewidth laser diodes and narrow linewidth laser diodes respectively. It is found that the MI threshold increases by 26% when the amplifier is pumped by the broad linewidth laser diodes. Experiment results show that the MI threshold is affected by the local heat load rather than the average or the total heat load. The calculation shows that the local heat deposit actually plays the key role to stimulate the MI behaviour. At the MI threshold position in the fiber, the local heat deposit also changes dramatically. The effect of the thermal conductivity on the MI threshold is also studied. Our investigation shows that the MI threshold increases from 1269 W to 1950 W when the thermal conductivity of the fiber amplifier is increased from 0.3 W/(m . K) to 5 W/(m . K). Through optimizing the pump linewidth and the cooling efficiency of the gain fiber, the MI threshold is doubled in our experiment. (copyright 2017 by WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. 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

  13. 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.

  14. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature; Comportements metallurqigue et mecanique des materiaux de gainage du combustible REP oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases ({alpha}(O) and 'ex {beta}') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the {beta}-->{alpha} phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials

  15. 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

  16. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  17. 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.

  18. 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

  19. 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.

  20. 3D FE simulation of PCMI (Pellet-Cladding Mechanical Interaction) considering frictionless contact

    International Nuclear Information System (INIS)

    Seo, Sang-Kyu; Lee, Sung-Uk; Lee, Eun-Ho; Yang, Dong-Yol; Kim, Hyo-Chan; Yang, Yong-Sik

    2014-01-01

    The goal of this code is coupling every aspect of physical phenomenon. Monodimensional FE model has been made for METEOR. It is good to evaluate the global behavior in high burn up levels. However, the multi-dimensional PCI analysis code is necessary to precisely analyze the stress distribution especially in case of the crack analysis. CAST3M 3D finite element code has been developed considering thermo-mechanical interaction in detail for TOUTATIS code. The advanced multidimensional code called ALCYONE has been developed considering chemical-physics and thermomechanical aspects. Although there are many codes that analyze pellet and cladding interaction, it is difficult to consider every physical aspect. In this paper, pellet to cladding mechanical interaction in 3D has been simulated with frictionless contact using the developed module, which is written in FORTRANN90. In this paper, 3D PCMI FE model is simulated with frictionless contact and elastic deformation. From the frictionless contact analysis, the interfacial pressure has been calculated and then this is used to obtain the solid heat coefficient which is a main factor to analyze the thermal distribution

  1. High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nelson, Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parkison, Adam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-12-23

    The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H2O with the alumina reaction tube at 1700°C resulted in Al(OH)3

  2. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Maeda, Koji [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2011-09-30

    Highlights: > We evaluated diametral strain of fast reactor MOX fuel pins irradiated to 130 GWd/t. > The strain was due to cladding void swelling and irradiation creep. > The irradiation creep was caused by internal gas pressure and PCMI. > The PCMI was associated with pellet swelling by rim structure or by cesium uranate. > The latter effect tended to increase the cumulative damage fraction of the cladding. - Abstract: The C3M irradiation test, which was conducted in the experimental fast reactor, 'Joyo', demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, 'Monju'. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and {sup 137}Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  3. Effect of high hydrogen content on metallurgical and mechanical properties of zirconium alloy claddings after heat-treatment at high temperature

    International Nuclear Information System (INIS)

    Turque, Isabelle

    2016-01-01

    Under hypothetical loss-of-coolant accident conditions, fuel cladding tubes made of zirconium alloys can be exposed to steam at high temperature (HT, up 1200 C) before being cooled and then quenched in water. In some conditions, after burst occurrence the cladding can rapidly absorb a significant amount of hydrogen (secondary hydriding), up to 3000 wt.ppm locally, during steam exposition at HT. The study deals with the effect, poorly studied up to date, of high contents of hydrogen on the metallurgical and mechanical properties of two zirconium alloys, Zircaloy-4 and M5, during and after cooling from high temperatures, at which zirconium is in its β phase. A specific facility was developed to homogeneously charge in hydrogen up to ∼ 3000 wt.ppm cladding tube samples of several centimeters in length. Phase transformations, chemical element partitioning and hydrogen precipitation during cooling from the β temperature domain of zirconium were studied by using several techniques, for the materials containing up to ∼ 3000 wt.ppm of hydrogen in average: in-situ neutron diffraction upon cooling from 700 C, X-ray diffraction, μ-ERDA, EPMA and electron microscopy in particular. The results were compared to thermodynamic predictions. In order to study the effect of high hydrogen contents on the mechanical behavior of the (prior-)μ phase of zirconium, axial tensile tests were performed at various temperatures between 20 and 700 C upon cooling from the β temperature domain, on samples with mean hydrogen contents up to ∼ 3000 wt.ppm. The results show that metallurgical and mechanical properties of the (prior-)β phase of zirconium alloys strongly depend on temperature and hydrogen content. (author) [fr

  4. Unirradiated cladding rip-propagation tests

    International Nuclear Information System (INIS)

    Hu, W.L.; Hunter, C.W.

    1981-04-01

    The size of cladding rips which develop when a fuel pin fails can affect the subassembly cooling and determine how rapidly fuel escapes from the pin. The object of the Cladding Rip Propagation Test (CRPT) was to quantify the failure development of cladding so that a more realistic fuel pin failure modeling may be performed. The test results for unirradiated 20% CS 316 stainless steel cladding show significantly different rip propagation behavior at different temperatures. At room temperature, the rip growth is stable as the rip extension increases monotonically with the applied deformation. At 500 0 C, the rip propagation becomes unstable after a short period of stable rip propagation. The rapid propagation rate is approximately 200 m/s, and the critical rip length is 9 mm. At test temperatures above 850 0 C, the cladding exhibits very high failure resistances, and failure occurs by multiple cracking at high cladding deformation. 13 figures

  5. Dry sliding wear behavior of laser clad TiVCrAlSi high entropy alloy coatings on Ti–6Al–4V substrate

    International Nuclear Information System (INIS)

    Huang, Can; Zhang, Yongzhong; Vilar, Rui; Shen, Jianyun

    2012-01-01

    Highlights: ► TiVCrAlSi high entropy alloy coatings were obtained on Ti–6Al–4V by laser cladding. ► (Ti,V) 5 Si 3 forms because the formation is accompanied of large variation on enthalpy. ► Wear resistance of Ti–6Al–4V is improved by laser cladding with TiVCrAlSi. ► The wear mechanism is investigated. -- Abstract: Approximately equimolar ratio TiVCrAlSi high entropy alloy coatings has been deposited by laser cladding on Ti–6Al–4V alloy. The analysis of the microstructure by scanning electron microscopy (SEM) shows that the coating is metallurgically bonded to the substrate. X-ray diffraction (XRD) and energy dispersive spectrometer (EDS) analyses show that TiVCrAlSi coating is composed of precipitates of (Ti,V) 5 Si 3 dispersed in a body-centered cubic (BCC) matrix. Intermetallic compound (Ti,V) 5 Si 3 forms because the formation is accompanied by larger variation on enthalpy, which may offset the entropy term. The dry sliding wear tests show that the wear resistance of Ti–6Al–4V is improved by laser cladding with TiVCrAlSi. The enhancement of the wear resistance is explained by the presence of the hard silicide phase dispersed in a relatively ductile BCC matrix, which allows sliding wear to occur in the mild oxidative regime for a wide range of testing conditions.

  6. Laser surface cladding:a literature survey

    OpenAIRE

    Gedda, Hans

    2000-01-01

    This work consists of a literature survey of a laser surface cladding in order to investigate techniques to improve the cladding rate for the process. The high local heat input caused by the high power density of the laser generates stresses and the process is consider as slow when large areas are processed. To avoid these disadvantages the laser cladding process velocity can be increased three or four times by use of preheated wire instead of the powder delivery system. If laser cladding is ...

  7. High-beam quality, high-efficiency laser based on fiber with heavily Yb(3+)-doped phosphate core and silica cladding.

    Science.gov (United States)

    Egorova, O N; Semjonov, S L; Medvedkov, O I; Astapovich, M S; Okhrimchuk, A G; Galagan, B I; Denker, B I; Sverchkov, S E; Dianov, E M

    2015-08-15

    We have fabricated and tested a composite fiber with an Yb(3+)-doped phosphate glass core and silica cladding. Oscillation with a slope efficiency of 74% was achieved using core pumping at 976 nm with fiber lengths of 48-90 mm in a simple laser configuration, where the cavity was formed by a high-reflectivity Bragg grating and the cleaved fiber end. The measured M(2) factors were as low as 1.05-1.22 even though the fiber was multimode at the lasing wavelength.

  8. High-power and highly efficient diode-cladding-pumped holmium-doped fluoride fiber laser operating at 2.94 microm.

    Science.gov (United States)

    Jackson, Stuart D

    2009-08-01

    A high-power diode-cladding-pumped Ho(3+), Pr(3+)-doped fluoride glass fiber laser is demonstrated. The laser produced a maximum output power of 2.5 W at a slope efficiency of 32% using diode lasers emitting at 1,150 nm. The long-emission wavelength of 2.94 microm measured at maximum pump power, which is particularly suited to medical applications, indicates that tailoring of the proportion of Pr(3+) ions can provide specific emission wavelengths while providing sufficient de-excitation of the lower laser level.

  9. Stress intensity factor at the tip of cladding incipient crack in RIA-simulating experiments for high-burnup PWR fuels

    International Nuclear Information System (INIS)

    Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi

    2009-01-01

    RIA-simulating experiments for high-burnup PWR fuels have been performed in the NSRR, and the stress intensity factor K 1 at the tip of cladding incipient crack has been evaluated in order to investigate its validity as a PCMI failure threshold under RIA conditions. An incipient crack depth was determined by observation of metallographs. The maximum hydride-rim thickness in the cladding of the test fuel rod was regarded as the incipient crack depth in each test case. Hoop stress in the cladding periphery during the pulse power transient was calculated by the RANNS code. K 1 was calculated based on crack depth and hoop stress. According to the RANNS calculation, PCMI failure cases can be divided into two groups: failure in the elastic phase and failure in the plastic phase. In the former case, elastic deformation was predominant around the incipient crack at failure time. K 1 is available only in this case. In the latter, plastic deformation was predominant around the incipient crack at failure time. Failure in the elastic phase never occurred when K 1 was less than 17 MPa m 1/2 . For failure in the plastic phase, the plastic hoop strain of the cladding periphery at failure time clearly showed a tendency to decrease with incipient crack depth. The combination of K 1 , for failure in the elastic phase, and plastic hoop strain at failure, for failure in the plastic phase, can be an effective index of PCMI failure under RIA conditions. (author)

  10. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  11. The adequacy of methods used for the approval of high burnup core loading

    International Nuclear Information System (INIS)

    Sonnenburg, H.G.

    2002-01-01

    New fuel assembly designs and new core loading strategies are foreseen by most utilities, optimising the use of nuclear fuel in nuclear power plants. Increasing the burn-up to high values above 50 MWd/kg affects the fuel and cladding conditions, which could have safety relevant consequences. It is the task of the safety authorities to assess the impact of these changes with respect to compliance with safety regulations. Usually this assessment is based on code analyses which contain models developed at a time when the burn-up was significantly lower. Because the high burn-up is accompanied with the development of new phenomena like the rim effect on fuel pellets, the codes' models need to be revised for the representation of these new phenomena. The objective of this paper is to present a review of the knowledge base of the fuel phenomena under high-burn-up conditions as seen from safety aspects. The safety relevant fuel rod phenomena will be discussed. It will further provide an assessment of the limitations of the methodologies so far applied in the context of LOCA and RIA transients. The recently started research activities in Germany to improve the methodologies will be presented. (author)

  12. 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

  13. Modelling cladding response to changing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville; Ikonen, Timo [VTT Technical Research Centre of Finland ltd (Finland)

    2016-11-15

    The cladding of the nuclear fuel is subjected to varying conditions during fuel reactor life. Load drops and reversals can be modelled by taking cladding viscoelastic behaviour into account. Viscoelastic contribution to the deformation of metals is usually considered small enough to be ignored, and in many applications it merely contributes to the primary part of the creep curve. With nuclear fuel cladding the high temperature and irradiation as well as the need to analyse the variable load all emphasise the need to also inspect the viscoelasticity of the cladding.

  14. Pulsed Laser Cladding of Ni Based Powder

    Science.gov (United States)

    Pascu, A.; Stanciu, E. M.; Croitoru, C.; Roata, I. C.; Tierean, M. H.

    2017-06-01

    The aim of this paper is to optimize the operational parameters and quality of one step Metco Inconel 718 atomized powder laser cladded tracks, deposited on AISI 316 stainless steel substrate by means of a 1064 nm high power pulsed laser, together with a Precitec cladding head manipulated by a CLOOS 7 axes robot. The optimization of parameters and cladding quality has been assessed through Taguchi interaction matrix and graphical output. The study demonstrates that very good cladded layers with low dilution and increased mechanical proprieties could be fabricated using low laser energy density by involving a pulsed laser.

  15. Corrosion fatigue crack growth in clad low-alloy steels. Part 2: Water flow rate effects in high-sulfur plate steel

    International Nuclear Information System (INIS)

    James, L.A.; Lee, H.B.; Wire, G.L.; Novak, S.R.; Cullen, W.H.

    1997-01-01

    Corrosion fatigue crack propagation tests were conducted on a high-sulfur ASTM A302-B plate steel overlaid with weld-deposited Alloy EN82H cladding. The specimens featured semi-elliptical surface cracks penetrating approximately 6.3 mm of cladding into the underlying steel. The initial crack sizes were relatively large with surface lengths of 22.8--27.3 mm, and depths of 10.5--14.1 mm. The experiments were initiated in a quasi-stagnant low-oxygen (O 2 < 10 ppb) aqueous environment at 243 C, under loading conditions (ΔK, R, cyclic frequency) conducive to environmentally assisted cracking (EAC) under quasi-stagnant conditions. Following fatigue testing under quasi-stagnant conditions where EAC was observed, the specimens were then fatigue tested under conditions where active water flow of either 1.7 m/s or 4.7 m/s was applied parallel to the crack. Earlier experiments on unclad surface-cracked specimens of the same steel exhibited EAC under quasi-stagnant conditions, but water flow rates at 1.7 m/s and 5.0 m/s parallel to the crack mitigated EAC. In the present experiments on clad specimens, water flow at approximately the same as the lower of these velocities did not mitigate EAC, and a free stream velocity approximately the same as the higher of these velocities resulted in sluggish mitigation of EAC. The lack of robust EAC mitigation was attributed to the greater crack surface roughness in the cladding interfering with flow induced within the crack cavity. An analysis employing the computational fluid dynamics code, FIDAP, confirmed that frictional forces associated with the cladding crack surface roughness reduced the interaction between the free stream and the crack cavity

  16. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

  17. Modelling of pellet cladding interaction during power ramps in PWR rods by means of Transuranus fuel rod analysis code

    International Nuclear Information System (INIS)

    Di Marcello, V.; Luzzi, L.

    2008-01-01

    Pellet-cladding interaction (PCI) in PWR type rods subjected to power ramps was analysed by means of TRANSURANUS (TU) fuel rod performance code. PCI phenomena depend on the fuel power history - i.e. by several irradiation and thermal induced phenomena occurring in the fuel rod and mutually interacting during its life in reactor - and may become critical for cladding integrity under accidental conditions. Ten test fuel rods, whose power histories and post irradiation experiment (PIE) data were available from the OECD/NEA-IAEA International Fuel Performance Experiment (UTE) database through the Studsvik SUPER-RAMP Project, were simulated by TRANSURANUS. During a power ramp pellet gaseous swelling can be inhibited by cladding pressure and can be over-predicted by a normal operation swelling model. This phenomenon was simulated by a new formulation of a fuel swelling model already available in the code, in order to consider hot pressing of inter-granular -fuel porosity due to the high hydrostatic stress resulting from PCI: it was found that TRANSURANUS, as a result of the proposed swelling formulation as well as of the accurate modelling of the other phenomena occurring during irradiation, gives correct predictions on PCI induced fuel rod failures. In addition, PCI failure threshold identified by TRANSURANUS was compared with the technological limits known in literature: the possibility of relaxing these limits for low burn-up values and the preponderance of the European fuel rod design in front of PCI emerged from TU analyses. Finally, a good agreement was found between TU evaluations and PIE data, with regard to fission gas release, fuel grain growth, and creep, corrosion and elongation of the cladding. (authors)

  18. 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

  19. Cladding using a 15 kW CO2 laser

    International Nuclear Information System (INIS)

    Vesely, E.J.; Verma, S.K.

    1989-01-01

    Laser alloying or cladding differs little in principle from the traditional forms of weld overlays, but lasers as a heat source offer some distinct advantages. With the selective heating attainable using high power lasers, good metallurgical bond of the clad layer, minimal dilution and typically, a very fine homogeneous microstructure can be obtained in the clad layer. This is a review of work in laser cladding using the 15 kW CO 2 laser. The authors discuss the ability of the laser clad surface to increase the high temperature oxidation resistance of a low-alloy carbon steel (4140). Examples of clads subjected to high- temperature thermal cycling of nickel-20% aluminum and TaC + 4140 clad low-alloy steel and straight high-temperature oxidation of Stellite 6-304L cladding on a 4140 substrate are given

  20. Characterization of mechanical properties and microstructure of highly irradiated SS 316

    Energy Technology Data Exchange (ETDEWEB)

    Karthik, V., E-mail: karthik@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kumar, RanVijay; Vijayaragavan, A.; Venkiteswaran, C.N.; Anandaraj, V.; Parameswaran, P.; Saroja, S.; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.; Jayakumar, T.; Raj, Baldev [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2013-08-15

    Cold worked austenitic stainless steel type AISI 316 is used as the material for fuel cladding and wrapper of the Fast Breeder Test Reactor (FBTR), India. The evaluation of mechanical properties of these core structurals is very essential to assess its integrity and ensure safe and productive operation of FBTR to very high burn-ups. The changes in the mechanical properties of these core structurals are associated with microstructural changes caused by high fluence neutron irradiation and temperatures of 673–823 K. Remote tensile testing has been used for evaluating the tensile properties of irradiated clad tubes and shear punch test using small disk specimens for evaluating the properties of irradiated hexagonal wrapper. This paper will highlight the methods employed for evaluating the mechanical properties of the irradiated cladding and wrapper and discuss the trends in properties as a function of dpa (displacement per atom) and irradiation temperature.

  1. Heat suppression of the fiber coating on a cladding light stripper in high-power fiber laser.

    Science.gov (United States)

    Yan, Ming-Jian; Wang, Zheng; Meng, Ling-Qiang; Yin, Lu; Han, Zhi-Gang; Shen, Hua; Wang, Hai-Lin; Zhu, Ri-Hong

    2018-01-20

    We present a theoretical model for the thermal effect of the fiber coating on a high-power cladding light stripper, which is fabricated by chemical etching. For the input and output of the fiber coating, a novel segmented corrosion method and increasing attenuation method are proposed for heat suppression, respectively. The relationship between the attenuation and temperature rise of the fiber coating at the output is experimentally demonstrated. The temperature distribution of the fiber coating at the input as well as the return light power caused by scattering are measured for the etched fiber with different surface roughness values. The results suggest that the rise in temperature is primarily caused by the scattering light propagating into the coating. Finally, an attenuation of 27 dB is achieved. At a room temperature of 23°C and input pump power of 438 W, the highest temperature of the input fiber coating decreases from 39.5°C to 27.9°C by segmented corrosion, and the temperature rise of the output fiber coating is close to 0.

  2. 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

  3. 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

  4. Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

    Science.gov (United States)

    Grisaffe, Salvatore J.; Caves, Robert M.

    1964-01-01

    An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

  5. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Walker, Larry R [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-06-01

    This report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for cladding large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high-power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for cladding inaccessible surfaces such as the interior surfaces of heat exchangers. An initial evaluation for performed on the quality of nickel claddings processed using the two selected cladding techniques.

  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. High-power blue laser diodes with indium tin oxide cladding on semipolar (202{sup ¯}1{sup ¯}) GaN substrates

    Energy Technology Data Exchange (ETDEWEB)

    Pourhashemi, A., E-mail: pourhashemi@engr.ucsb.edu; Farrell, R. M.; Cohen, D. A.; Speck, J. S. [Materials Department, University of California, Santa Barbara, California 93106 (United States); DenBaars, S. P.; Nakamura, S. [Materials Department, University of California, Santa Barbara, California 93106 (United States); Department of Electrical and Computer Engineering, University of California, Santa Barbara, California 93106 (United States)

    2015-03-16

    We demonstrate a high power blue laser diode (LD) using indium tin oxide as a cladding layer on semipolar oriented GaN. These devices show peak output powers and external quantum efficiencies comparable to state-of-the-art commercial c-plane devices. Ridge waveguide LDs were fabricated on (202{sup ¯}1{sup ¯}) oriented GaN substrates using InGaN waveguiding layers and GaN cladding layers. At a lasing wavelength of 451 nm at room temperature, an output power of 2.52 W and an external quantum efficiency of 39% were measured from a single facet under a pulsed injection current of 2.34 A. The measured differential quantum efficiency was 50%.

  8. Evaluation of Thermal Creep and Hydride Re-orientation Properties of High Burnup Spent Fuel Cladding under Long Term Dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Kamimura, K [JNES (Japan)

    2012-07-01

    In Japan, spent fuels will be reprocessed as recyclable energy source at a reprocessing plant. The first commercial plant is under-constructing and will start operation in 2008. It is necessary that spent fuels should be stored in the independent interim storage facilities (ISF) until reprocessing. Utilities plan the operation of the first ISF in 2010. JNES has a mission to support the safety body by researching the data of technical standard and regulation. Investigating of spent fuel integrity during long term dry storage is one of them. The objectives are: 1) Evaluation of the effects of material design changes on creep properties of high burnup spent fuel cladding; 2) Evaluation of the effects of alloy elements and texture of irradiated Zircaloy on hydride re-orientation properties and the effects of radial hydrides on cladding mechanical properties; 3) Evaluation of the effects of temperature on irradiation hardening recovery.

  9. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

    2015-01-01

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  10. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  11. Polarization effects in silicon-clad optical waveguides

    Science.gov (United States)

    Carson, R. F.; Batchman, T. E.

    1984-01-01

    By changing the thickness of a semiconductor cladding layer deposited on a planar dielectric waveguide, the TE or TM propagating modes may be selectively attenuated. This polarization effect is due to the periodic coupling between the lossless propagating modes of the dielectric slab waveguide and the lossy modes of the cladding layer. Experimental tests involving silicon claddings show high selectivity for either polarization.

  12. Folded cladding porous shaped photonic crystal fiber with high sensitivity in optical sensing applications: Design and analysis

    Directory of Open Access Journals (Sweden)

    Bikash Kumar Paul

    2017-02-01

    Full Text Available A micro structure folded cladding porous shaped with circular air hole photonic crystal fiber (FP-PCF is proposed and numerically investigated in a broader wavelength range from 1.4 µm to 1.64 µm (E+S+C+L+U for chemical sensing purposes. Employing finite element method (FEM with anisotropic perfectly matched layer (PML various properties of the proposed FP-PCF are numerically inquired. Filling the hole of core with aqueous analyte ethanol (n = 1.354 and tuning different geometric parameters of the fiber, the sensitivity order of 64.19% and the confinement loss of 2.07 × 10-5 dB/m are attained at 1.48 µm wavelength in S band. The investigated numerical simulation result strongly focuses on sensing purposes; because this fiber attained higher sensitivity with lower confinement loss over the operating wavelength. Measuring time of sensitivity, simultaneously confinement loss also inquired. It reflects that confinement loss is highly dependable on PML depth but not for sensitivity. Beside above properties numerical aperture (NA, nonlinearity, and effective area are also computed. This FP-PCF also performed as sensor for other alcohol series (methanol, propanol, butanol, pentanol. Optimized FP-PCF shows higher sensitivity and low confinement loss carrying high impact in the area of chemical as well as gas sensing purposes. Surely it is clear that install such type of sensor will flourish technology massively.         Keywords: Confinement loss, Effective area, Index guiding FP-PCF, Numerical aperture, Nonlinear coefficient, Sensitivity

  13. Characterization of the Microstructure in Recrystallized Zircaloy-2 Cladding Irradiated to a High Neutron Dose

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2003-04-01

    The objectives of the present project were to determine if there is anything in the microstructure of highly irradiated Zircaloy-2 which may make the material fracture in a brittle manner. Samples were taken from three different locations on a fuel rod which had been irradiated for 12 years. The displacement doses were estimated to be 1.4, 9 and 28 dpa. Specimens for electron microscopy were prepared with two different orientation called axial and radial. In the axial orientation the electron beam goes parallel with the basal plane and diffraction conditions can be arranged so that dislocations with a Burgers' vectors become invisible. In the low dose specimen only a-component damage was present and all second phase particles were crystalline. In both the high and intermediate dose samples there was c-component damage present with a slightly higher amount in the high dose sample. The particles of the Zr(Cr,Fe) 2 type were generally amorphous in these samples and the Fe-content of the particles was highly reduced. The hydride structures were similar in all samples. The hydrides were often precipitated in parallel in the same grain and chains of hydrides were seen which ran from grain to grain. No population of small hydrides were observed except from surface hydrides formed during specimen preparation. It was concluded from the investigation that there is nothing in the microstructure which may make the material in the high dose state subject to a purely mechanically induced fast brittle cracking

  14. 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)

  15. Comparison of corrosion behavior between fusion cladded and explosive cladded Inconel 625/plain carbon steel bimetal plates

    International Nuclear Information System (INIS)

    Zareie Rajani, H.R.; Akbari Mousavi, S.A.A.; Madani Sani, F.

    2013-01-01

    Highlights: ► Both explosive and fusion cladding aggravate the corrosion resistance of Inconel 625. ► Fusion cladding is more detrimental to nonuniform corrosion resistance. ► Single-layered fusion coat does not show any repassivation ability. ► Adding more layers enhance the corrosion resistance of fusion cladding Inconel 625. ► High impact energy spoils the corrosion resistance of explosive cladding Inconel 625. -- Abstract: One of the main concerns in cladding Inconel 625 superalloy on desired substrates is deterioration of corrosion resistance due to cladding process. The present study aims to compare the effect of fusion cladding and explosive cladding procedures on corrosion behavior of Inconel 625 cladding on plain carbon steel as substrate. Also, an attempt has been made to investigate the role of load ratio and numbers of fusion layers in corrosion behavior of explosive and fusion cladding Inconel 625 respectively. In all cases, the cyclic polarization as an electrochemical method has been applied to assess the corrosion behavior. According to the obtained results, both cladding methods aggravate the corrosion resistance of Inconel 625. However, the fusion cladding process is more detrimental to nonuniform corrosion resistance, where the chemical nonuniformity of fusion cladding superalloy issuing from microsegregation, development of secondary phases and contamination of clad through dilution hinders formation of a stable passive layer. Moreover, it is observed that adding more fusion layers can enhance the nonuniform corrosion resistance of fusion cladding Inconel 625, though this resistance still remains weaker than explosive cladding superalloy. Also, the results indicate that raising the impact energy in explosive cladding procedure drops the corrosion resistance of Inconel 625.

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. 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)

  1. Microstructure and wear behaviors of laser clad NiCr/Cr3C2-WS2 high temperature self-lubricating wear-resistant composite coating

    Science.gov (United States)

    Yang, Mao-Sheng; Liu, Xiu-Bo; Fan, Ji-Wei; He, Xiang-Ming; Shi, Shi-Hong; Fu, Ge-Yan; Wang, Ming-Di; Chen, Shu-Fa

    2012-02-01

    The high temperature self-lubricating wear-resistant NiCr/Cr3C2-30%WS2 coating and wear-resistant NiCr/Cr3C2 coating were fabricated on 0Cr18Ni9 austenitic stainless steel by laser cladding. Phase constitutions and microstructures were investigated, and the tribological properties were evaluated using a ball-on-disc wear tester under dry sliding condition at room-temperature (17 °C), 300 °C and 600 °C, respectively. Results indicated that the laser clad NiCr/Cr3C2 coating consisted of Cr7C3 primary phase and γ-(Fe,Ni)/Cr7C3 eutectic colony, while the coating added with WS2 was mainly composed of Cr7C3 and (Cr,W)C carbides, with the lubricating WS2 and CrS sulfides as the minor phases. The wear tests showed that the friction coefficients of two coatings both decrease with the increasing temperature, while the both wear rates increase. The friction coefficient of laser clad NiCr/Cr3C2-30%WS2 is lower than the coating without WS2 whatever at room-temperature, 300 °C, 600 °C, but its wear rate is only lower at 300 °C. It is considered that the laser clad NiCr/Cr3C2-30%WS2 composite coating has good combination of anti-wear and friction-reducing capabilities at room-temperature up to 300 °C.

  2. Analysis of corrosion behavior of KOFA cladding

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, Ki Hang; Seo, Keum Seok; Chung, Jin Gon

    1994-01-01

    The corrosion behavior of KOFA cladding was analyzed using the oxide measurement data of KOFA fuel irradiated up to the fuel rod burnup of 35,000 MWD/MTU for two cycles in Kori-2. Even though KOFA cladding is a standard Zircaloy-4 manufactured by Westinghouse according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification, it was expected that in-pile corrosion behavior of KOFA cladding would not be equivalent to that of Siemens/KWU's cladding due to the differences in such manufacturing processes as cold work and heat treatment. The analysis of measured KOFA cladding oxidation showed that oxidation of KOFA cladding is at least 19 % lower than the design analysis based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Lower corrosion of KOFA cladding seems to result from the differences in the manufacturing processes and chemical composition although the burnup and oxide layer thickness of the measured fuel rods is relatively low and the amount of the oxidation data base is small

  3. High pumping-power fiber combiner for double-cladding fiber lasers and amplifiers

    Science.gov (United States)

    Zheng, Jinkun; Zhao, Wei; Zhao, Baoyin; Li, Zhe; Chang, Chang; Li, Gang; Gao, Qi; Ju, Pei; Gao, Wei; She, Shengfei; Wu, Peng; Hou, Chaoqi; Li, Weinan

    2018-03-01

    A high pumping-power fiber combiner for backward pumping configurations is fabricated and demonstrated by manufacturing process refinement. The pump power handling capability of every pump fiber can extend to 600 W, corresponding to the average pump coupling efficiency of 94.83%. Totally, 2.67-kW output power with the beam quality factor M2 of 1.41 was obtained, using this combiner in the fiber amplifier experimental setup. In addition, the temperature of the splicing region was less than 50.0°C in the designed combiner under the action of circulating cooling water. The experimental results prove that the designed combiner is a promising integrated all-fiber device for multikilowatt continuous-wave fiber laser with excellent beam quality.

  4. Cladding tube manufacturing technology

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report A lloy Development for High Burnup Cladding . Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs

  5. Cladding tube manufacturing technology

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report 'Alloy Development for High Burnup Cladding.' Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs.

  6. 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)

  7. 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.)

  8. 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

  9. 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.

  10. Resonantly cladding-pumped Yb-free Er-doped LMA fiber laser with record high power and efficiency.

    Science.gov (United States)

    Zhang, Jun; Fromzel, Viktor; Dubinskii, Mark

    2011-03-14

    We report the results of our power scaling experiments with resonantly cladding-pumped Er-doped eye-safe large mode area (LMA) fiber laser. While using commercial off-the-shelf LMA fiber we achieved over 88 W of continuous-wave (CW) single transverse mode power at ~1590 nm while pumping at 1532.5 nm. Maximum observed optical-to-optical efficiency was 69%. This result presents, to the best of our knowledge, the highest power reported from resonantly-pumped Yb-free Er-doped LMA fiber laser, as well as the highest efficiency ever reported for any cladding-pumped Er-doped laser, either Yb-co-doped or Yb-free.

  11. 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

  12. History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

    Energy Technology Data Exchange (ETDEWEB)

    Larry Zirker; Nathan Jerred; Dr. Indrajit Charit; James Cole

    2012-03-01

    Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

  13. History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

    International Nuclear Information System (INIS)

    Zirker, Larry; Jerred, Nathan; Charit, Indrajit; Cole, James

    2012-01-01

    Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

  14. High-temperature irradiation of niobium-1 w/o zirconium-clad UO/sub 2/. [Compatibility with lithium

    Energy Technology Data Exchange (ETDEWEB)

    Kangilaski, M.; Fromm, E.O.; Lozier, D.H.; Storhok, V.W.; Gates, J.E.

    1965-06-28

    Twenty-four 0.225-in.-diameter and six 0.290-in.-diameter UO/sub 2/ specimens clad with 80 mils of niobium-1 w/o zirconium were irradiated to burnups of 1.4 to 6.0 at. % of uranium at surface temperatures of 900 to 1400/sup 0/C. UO/sub 2/ and lithium were found to be incompatible at these temperatures, and the thick cladding was used primarily to minimize the chances of contact of UO/sub 2/ and the lithium coolant. The thickly clad specimens did not undergo any dimensional changes as a result of irradiation, although it was found that movement of UO/sub 2/ took place in the axial direction by a vaporization-redeposition mechanism. It was found that 32 to 87% of the fission gases was released from the fuel, depending on the temperature of the specimen. Metallographic examination of longitudinal and transverse sections of the specimens indicated the usual UO/sub 2/ microstructure with columnar grains. Grain-boundary thickening was observed in the UO/sub 2/ at higher burnups. The oxygen/uranium ratio of UO/sub 2/ increased with increasing burnup.

  15. Laser cladding of stainless steel with a copper-silver alloy to generate surfaces of high antimicrobial activity

    Science.gov (United States)

    Hans, Michael; Támara, Juan Carlos; Mathews, Salima; Bax, Benjamin; Hegetschweiler, Andreas; Kautenburger, Ralf; Solioz, Marc; Mücklich, Frank

    2014-11-01

    Copper and silver are used as antimicrobial agents in the healthcare sector in an effort to curb infections caused by bacteria resistant to multiple antibiotics. While the bactericidal potential of copper and silver alone are well documented, not much is known about the antimicrobial properties of copper-silver alloys. This study focuses on the antibacterial activity and material aspects of a copper-silver model alloy with 10 wt% Ag. The alloy was generated as a coating with controlled intermixing of copper and silver on stainless steel by a laser cladding process. The microstructure of the clad was found to be two-phased and in thermal equilibrium with minor Cu2O inclusions. Ion release and killing of Escherichia coli under wet conditions were assessed with the alloy, pure silver, pure copper and stainless steel. It was found that the copper-silver alloy, compared to the pure elements, exhibited enhanced killing of E. coli, which correlated with an up to 28-fold increased release of copper ions. The results show that laser cladding with copper and silver allows the generation of surfaces with enhanced antimicrobial properties. The process is particularly attractive since it can be applied to existing surfaces.

  16. Modification of tribology and high-temperature behavior of Ti-48Al-2Cr-2Nb intermetallic alloy by laser cladding

    International Nuclear Information System (INIS)

    Liu Xiubo; Wang Huaming

    2006-01-01

    In order to improve the tribology and high-temperature oxidation properties of the Ti-48Al-2Cr-2Nb intermetallic alloy simultaneously, mixed NiCr-Cr 3 C 2 precursor powders had been investigated for laser cladding treatment to modify wear and high-temperature oxidation resistance of the material. The alloy samples were pre-placed with NiCr-80, 50 and 20%Cr 3 C 2 (wt.%), respectively, and laser treated at the same parameters, i.e., laser output power 2.8 kW, beam scanning speed 2.0 mm/s, beam dimension 1 mm x 18 mm. The treated samples underwent tests of microhardness, wear and high-temperature oxidation. The results showed that laser cladding with different constitution of mixed precursor NiCr-Cr 3 C 2 powders improved surface hardness in all cases. Laser cladding with NiCr-50%Cr 3 C 2 resulted in the best modification of tribology and high-temperature oxidation behavior. X-ray diffraction (XRD), optical microscope (OM), scanning electron microscopy (SEM) and energy-dispersive spectrometer (EDS) analyses indicated that the formation of reinforced Cr 7 C 3 , TiC and both continuous and dense Al 2 O 3 , Cr 2 O 3 oxide scales were supposed to be responsible for the modification of the relevant properties. As a result, the present work had laid beneficial surface engineering foundation for TiAl alloy applied as future light weight and high-temperature structural candidate materials

  17. High performance fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    {omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

  18. Micro-Structures and High-Temperature Friction-Wear Performances of Laser Cladded Cr–Ni Coatings

    Directory of Open Access Journals (Sweden)

    Li Jiahong

    2018-01-01

    Full Text Available Cr–Ni coatings with the mass ratios of 17% Cr–83% Ni, 20% Cr–80% Ni and 24% Cr–76% Ni were fabricated on H13 hot work mould steel using a laser cladding (LC. The surface–interface morphologies, chemical elements, surface roughness and phase composition of the obtained Cr–Ni coatings were analysed using a scanning electron microscope (SEM, energy disperse spectroscopy (EDS, atomic force microscope (AFM and X–ray diffractometer (XRD, respectively. The friction–wear properties and wear rates of Cr–Ni coatings with the different mass ratios of Cr and Ni at 600 °C were investigated, and the worn morphologies and wear mechanism of Cr–Ni coatings were analysed. The results show that the phases of Cr–Ni coatings with mass ratios of 17% Cr–83% Ni, 20% Cr–80% Ni and 24% Cr–76% Ni are composed of Cr + Ni single-phases and their compounds at the different stoichiometry, the porosities on the Cr–Ni coatings increase with the Cr content increasing. The average coefficient of friction (COF of 17% Cr–83% Ni, 20% Cr–80% Ni and 24% Cr–76% coatings are 1.10, 0.33 and 0.87, respectively, in which the average COF of 20% Cr–80% Ni coating is the lowest, exhibiting the better anti-friction performance. The wear rate of 17% Cr–83% Ni, 20% Cr–80% Ni and 24% Cr–76% Ni coatings is 4.533 × 10−6, 5.433 × 10−6, and 1.761 × 10−6 N−1·s−1, respectively, showing the wear resistance of Cr–Ni coatings at a high temperature increases with the Cr content, in which the wear rate is 24% Cr–76% Ni coating with the better reducing wear. The wear mechanism of 17% Cr–83% Ni and 20% Cr–80% Ni and 24% Cr–76% coatings at 600 °C is primarily adhesive wear, and that of 24% Cr–76% coating is also accompanied by oxidative wear.

  19. Laser cladding with powder

    NARCIS (Netherlands)

    Schneider, M.F.; Schneider, Marcel Fredrik

    1998-01-01

    This thesis is directed to laser cladding with powder and a CO2 laser as heat source. The laser beam intensity profile turned out to be an important pa6 Summary rameter in laser cladding. A numerical model was developed that allows the prediction of the surface temperature distribution that is

  20. FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

    Directory of Open Access Journals (Sweden)

    WEON-JU KIM

    2013-08-01

    Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

  1. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  2. High-efficiency ytterbium-free erbium-doped all-glass double cladding silicate glass fiber for resonantly-pumped fiber lasers.

    Science.gov (United States)

    Qiang, Zexuan; Geng, Jihong; Luo, Tao; Zhang, Jun; Jiang, Shibin

    2014-02-01

    A highly efficient ytterbium-free erbium-doped silicate glass fiber has been developed for high-power fiber laser applications at an eye-safe wavelength near 1.55 μm. Our preliminary experiments show that high laser efficiency can be obtained from a relatively short length of the gain fiber when resonantly pumped at 1535 nm in both core- and cladding-pumping configurations. With a core-pumping configuration as high as 75%, optical-to-optical efficiency and 4 W output power were obtained at 1560 nm from a 1 m long gain fiber. When using a cladding-pumping configuration, approximately 13 W output power with 67.7% slope efficiency was demonstrated from a piece of 2 m long fiber. The lengths of silicate-based gain fiber are much shorter than their silica-based counterparts used in other experiments, which is significantly important for high-power narrow-band and/or pulsed laser applications.

  3. High-power linearly-polarized operation of a cladding-pumped Yb fibre laser using a volume Bragg grating for wavelength selection.

    Science.gov (United States)

    Jelger, P; Wang, P; Sahu, J K; Laurell, F; Clarkson, W A

    2008-06-23

    In this work a volume Bragg grating is used as a wavelength selective element in a high-power cladding-pumped Yb-doped silica fiber laser. The laser produced 138 W of linearly-polarized single-spatial-mode output at 1066 nm with a relatively narrow linewidth of 0.2 nm for approximately 202 W of launched pump power at 976 nm. The beam propagation factor (M(2)) for the output beam was determined to be 1.07. Thermal limitations of volume Bragg gratings are discussed in the context of power scaling for fiber lasers.

  4. Research on laser cladding control system based on fuzzy PID

    Science.gov (United States)

    Zhang, Chuanwei; Yu, Zhengyang

    2017-12-01

    Laser cladding technology has a high demand for control system, and the domestic laser cladding control system mostly uses the traditional PID control algorithm. Therefore, the laser cladding control system has a lot of room for improvement. This feature is suitable for laser cladding technology, Based on fuzzy PID three closed-loop control system, and compared with the conventional PID; At the same time, the laser cladding experiment and friction and wear experiment were carried out under the premise of ensuring the reasonable control system. Experiments show that compared with the conventional PID algorithm in fuzzy the PID algorithm under the surface of the cladding layer is more smooth, the surface roughness increases, and the wear resistance of the cladding layer is also enhanced.

  5. 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

  6. 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

  7. A high-throughput investigation of Fe-Cr-Al as a novel high-temperature coating for nuclear cladding materials.

    Science.gov (United States)

    Bunn, Jonathan Kenneth; Fang, Randy L; Albing, Mark R; Mehta, Apurva; Kramer, Matthew J; Besser, Matthew F; Hattrick-Simpers, Jason R

    2015-07-10

    High-temperature alloy coatings that can resist oxidation are urgently needed as nuclear cladding materials to mitigate the danger of hydrogen explosions during meltdown. Here we apply a combination of computationally guided materials synthesis, high-throughput structural characterization and data analysis tools to investigate the feasibility of coatings from the Fe–Cr–Al alloy system. Composition-spread samples were synthesized to cover the region of the phase diagram previous bulk studies have identified as forming protective oxides. The metallurgical and oxide phase evolution were studied via in situ synchrotron glancing incidence x-ray diffraction at temperatures up to 690 K. A composition region with an Al concentration greater than 3.08 at%, and between 20.0 at% and 32.9 at% Cr showed the least overall oxide growth. Subsequently, a series of samples were deposited on stubs and their oxidation behavior at 1373 K was observed. The continued presence of a passivating oxide was confirmed in this region over a period of 6 h.

  8. Characteristics of Ni-based coating layer formed by laser and plasma cladding processes

    International Nuclear Information System (INIS)

    Xu Guojian; Kutsuna, Muneharu; Liu Zhongjie; Zhang Hong

    2006-01-01

    The clad layers of Ni-based alloy were deposited on the SUS316L stainless plates by CO 2 laser and plasma cladding processes. The smooth clad bead was obtained by CO 2 laser cladding process. The phases of clad layer were investigated by an optical microscope, scanning electron microscopy (SEM), X-ray diffractometer (XRD), electron probe microanalysis (EPMA) and energy-dispersive spectrometer (EDS). The microstructures of clad layers belonged to a hypereutectic structure. Primary phases consist of boride CrB and carbide Cr 7 C 3 . The eutectic structure consists of Ni + CrB or Ni + Cr 7 C 3 . Compared with the plasma cladding, the fine microstructures, low dilutions, high Vickers hardness and excellent wear resistance were obtained by CO 2 laser cladding. All that show the laser cladding process has a higher efficiency and good cladding quality

  9. Out-of-pile experiments on the high-temperature behavior of Zircaloy-4 clad fuel rods

    International Nuclear Information System (INIS)

    Hagen, S.

    1984-01-01

    Out-of-pile experiments have been performed to investigate the escalation in temperature of Zircaloy-clad fuel rods during heatup in steam due to the exothermal Zircaloy steam reaction. In these tests single Zircaloy/uranium dioxide (UO 2 ) fuel rod simulators surrounded with a Zircaloy shroud--simulating the Zircaloy of neighboring rods--were heated inside a fiber ceramic insulation. The initial heating rates were varied from 0.3 to 2.5 K/s. In every test an escalation of the temperature rise rate was observed. The maximum measured surface temperature was about 2200 0 C. The temperature decreased after the maximum had been reached without decreasing the input electric power. The temperature decreases were due to inherent processes including the runoff of molten Zircaloy. The escalation process was influenced by the temperature behavior of the shroud, which was itself affected by the insulation and steam cooling. Damage to the fuel rods increased with increasing heatup rate. Fro slow heatup rates nearly no interaction between the oxidized cladding and UO 2 was observed, while for fast heatup rates the entire annular pellet was dissolved by molten Zircaloy

  10. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, N.; Horvath, R.; Pedersen, H.C.

    2005-01-01

    The present paper deals with the optimization of metal-clad waveguides for sensor applications to achieve high sensitivity for adlayer and refractive index measurements. By using the Fresnel reflection coefficients both the angular shift and the width of the resonances in the sensorgrams are taken...... into account. Our optimization shows that it is possible for metal-clad waveguides to achieve a sensitivity improvement of 600% compared to surface-plasmon-resonance sensors....

  11. 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

  12. Corrosion behaviour of cladded nickel base alloys

    International Nuclear Information System (INIS)

    Brandl, W.; Ruczinski, D.; Nolde, M.; Blum, J.

    1995-01-01

    As a consequence of the high cost of nickel base alloys their use as surface layers is convenient. In this paper the properties of SA-as well as RES-cladded NiMo 16Cr16Ti and NiCr21Mo14W being produced in single and multi-layer technique are compared and discussed with respect to their corrosion behaviour. Decisive criteria describing the qualities of the claddings are the mass loss, the susceptibility against intergranular corrosion and the pitting corrosion resistance. The results prove that RES cladding is the most suitable technique to produce corrosion resistant nickel base coatings. The corrosion behaviour of a two-layer RES deposition shows a better resistance against pitting than a three layer SAW cladding. 7 refs

  13. Stone cladding engineering

    National Research Council Canada - National Science Library

    Camposinhos, Rui de Sousa

    2014-01-01

    .... Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements...

  14. 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)

  15. Influences of Cr content and PWHT on microstructure and oxidation behavior of stainless steel weld overlay cladding materials in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Cao, X.Y.; Ding, X.F. [National Center for Materials Service Safety, University of Science and Technology Beijing, 30 Xueyuan Road, 100083 Beijing (China); Lu, Y.H., E-mail: lu_yonghao@mater.ustb.edu.cn [National Center for Materials Service Safety, University of Science and Technology Beijing, 30 Xueyuan Road, 100083 Beijing (China); Zhu, P. [Suzhou Nuclear Power Research Institute Co. Ltd., 1788 Xihuan Road, 215004 Suzhou (China); Shoji, T. [National Center for Materials Service Safety, University of Science and Technology Beijing, 30 Xueyuan Road, 100083 Beijing (China); Fracture and Reliability Research Institute, Tohoku University, 6-6-01 Aramaki Aoba, Aoba-ku, Sendai City 980-8579 (Japan)

    2015-12-15

    Influences of Cr content and post weld heat treatment (PWHT) on microstructure and oxidation behavior of stainless steel cladding materials in high temperature water were investigated. The amounts of metal oxidized and dissolved were estimated to compare the oxidation behaviors of cladding materials with different Cr contents and PWHT. The results indicated that higher Cr content led to formation of more ferrite content, and carbides were found along δ/γ phase interface after PWHT. Higher Cr content enhanced the pitting resistance and compactness of the oxide film to reduce metal amount oxidized and dissolved, which mitigated the weight changes and the formation of Fe-rich oxides. PWHT promoted more and deeper pitting holes along the δ/γ phase interface due to formation of carbides, which resulted in an increase in metal amount oxidized and dissolved, and were also responsible for more Fe-rich oxides and higher weight changes. - Highlights: • The amounts of metal oxidized and metal dissolved were estimated. • Higher Cr content increased ferrite content and PWHT led to formation of carbides. • PWHT promoted more and deeper pitting holes along the δ/γ phase interface. • Lower Cr content and PWHT promoted the metal amounts oxidized and dissolved. • Lower Cr content and PWHT increased weight changes and Fe-rich film formation.

  16. Capturing reflected cladding modes from a fiber Bragg grating with a double-clad fiber coupler.

    Science.gov (United States)

    Baiad, Mohamad Diaa; Gagné, Mathieu; Lemire-Renaud, Simon; De Montigny, Etienne; Madore, Wendy-Julie; Godbout, Nicolas; Boudoux, Caroline; Kashyap, Raman

    2013-03-25

    We present a novel measurement scheme using a double-clad fiber coupler (DCFC) and a fiber Bragg grating (FBG) to resolve cladding modes. Direct measurement of the optical spectra and power in the cladding modes is obtained through the use of a specially designed DCFC spliced to a highly reflective FBG written into slightly etched standard photosensitive single mode fiber to match the inner cladding diameter of the DCFC. The DCFC is made by tapering and fusing two double-clad fibers (DCF) together. The device is capable of capturing backward propagating low and high order cladding modes simply and efficiently. Also, we demonstrate the capability of such a device to measure the surrounding refractive index (SRI) with an extremely high sensitivity of 69.769 ± 0.035 μW/RIU and a resolution of 1.433 × 10(-5) ± 8 × 10(-9) RIU between 1.37 and 1.45 RIU. The device provides a large SRI operating range from 1.30 to 1.45 RIU with sufficient discrimination for all individual captured cladding modes. The proposed scheme can be adapted to many different types of bend, temperature, refractive index and other evanescent wave based sensors.

  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. Analyses on Silicide Coating for LOCA Resistant Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

  1. Analyses on Silicide Coating for LOCA Resistant Cladding

    International Nuclear Information System (INIS)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

    2015-01-01

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

  2. Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

    2016-08-15

    Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

  3. Laser cladding of turbine blades

    International Nuclear Information System (INIS)

    Shepeleva, L.; Medres, B.; Kaplan, W.D.; Bamberger, M.

    2000-01-01

    A comparative study of two different techniques for the application of wear-resistant coatings for contact surfaces of shroud shelves of gas turbine engine blades (GTE) has been conducted. Wear-resistant coatings were applied on In713 by laser cladding with direct injection of the cladding powder into the melt pool. Laser cladding was conducted with a TRUMPF-2500, CW-CO 2 laser. The laser cladding was compared with commercially available plasma cladding with wire. Both plasma and laser cladded zones were characterized by optical and scanning electron microscopy. It was found that the laser cladded zone has a higher microhardness value (650-820 HV) compared with that of the plasma treated material (420-440 HV). This is a result of the significant reduction in grain size in the case of laser cladding. Unlike the plasma cladded zones, the laser treated material is free of micropores and microcracks. (orig.)

  4. Laser cladding technology to small diameter pipes

    International Nuclear Information System (INIS)

    Fujimagari, H.; Hagiwara, M.; Kojima, T.

    2000-01-01

    A laser cladding method which produces a highly corrosion-resistant material coating layers (cladding) on the austenitic stainless steel (type 304 SS) pipe inner surface was developed to prevent SCC (stress corrosion cracking) occurrence. This technology is applicable to a narrow and long distance area from operators, because of the good accessibility of the YAG (yttrium-aluminum-garnet) laser beam that can be transmitted through an optical fiber. In this method a mixed paste metallic powder and heating-resistive organic solvent are firstly placed on the inner surface of a small pipe, and then a YAG laser beam transmitted through an optical fiber irradiates to the pasted area. A mixed paste will be melted and form a cladding layer subsequently. A cladding layer shows as excellent corrosion resistance property. This laser cladding (LC) method had already applied to several domestic nuclear power plants and had obtained a good reputation. This report introduces the outline of laser cladding technology, the developed equipment for practical application in the field, and the circumstance in actual plant application. (orig.)

  5. Initial Cladding Condition

    International Nuclear Information System (INIS)

    Siegmann, E.

    2000-01-01

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  6. 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)

  7. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    International Nuclear Information System (INIS)

    R. Schreiner

    2004-01-01

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database

  8. Phosphate-core silica-clad Er/Yb-doped optical fiber and cladding pumped laser.

    Science.gov (United States)

    Egorova, O N; Semjonov, S L; Velmiskin, V V; Yatsenko, Yu P; Sverchkov, S E; Galagan, B I; Denker, B I; Dianov, E M

    2014-04-07

    We present a composite optical fiber with a Er/Yb co-doped phosphate-glass core in a silica glass cladding as well as cladding pumped laser. The fabrication process, optical properties, and lasing parameters are described. The slope efficiency under 980 nm cladding pumping reached 39% with respect to the absorbed pump power and 28% with respect to the coupled pump power. Due to high doping level of the phosphate core optimal length was several times shorter than that of silica core fibers.

  9. Microstructure and Mechanical Properties of Laser Clad and Post-cladding Tempered AISI H13 Tool Steel

    Science.gov (United States)

    Telasang, Gururaj; Dutta Majumdar, Jyotsna; Wasekar, Nitin; Padmanabham, G.; Manna, Indranil

    2015-05-01

    This study reports a detailed investigation of the microstructure and mechanical properties (wear resistance and tensile strength) of hardened and tempered AISI H13 tool steel substrate following laser cladding with AISI H13 tool steel powder in as-clad and after post-cladding conventional bulk isothermal tempering [at 823 K (550 °C) for 2 hours] heat treatment. Laser cladding was carried out on AISI H13 tool steel substrate using a 6 kW continuous wave diode laser coupled with fiber delivering an energy density of 133 J/mm2 and equipped with a co-axial powder feeding nozzle capable of feeding powder at the rate of 13.3 × 10-3 g/mm2. Laser clad zone comprises martensite, retained austenite, and carbides, and measures an average hardness of 600 to 650 VHN. Subsequent isothermal tempering converted the microstructure into one with tempered martensite and uniform dispersion of carbides with a hardness of 550 to 650 VHN. Interestingly, laser cladding introduced residual compressive stress of 670 ± 15 MPa, which reduces to 580 ± 20 MPa following isothermal tempering. Micro-tensile testing with specimens machined from the clad zone across or transverse to cladding direction showed high strength but failure in brittle mode. On the other hand, similar testing with samples sectioned from the clad zone parallel or longitudinal to the direction of laser cladding prior to and after post-cladding tempering recorded lower strength but ductile failure with 4.7 and 8 pct elongation, respectively. Wear resistance of the laser surface clad and post-cladding tempered samples (evaluated by fretting wear testing) registered superior performance as compared to that of conventional hardened and tempered AISI H13 tool steel.

  10. Impact of reactor water chemistry on cladding performance

    International Nuclear Information System (INIS)

    Cox, B.

    1997-01-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  11. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B. [University of Toronto, Centre for Nuclear Engineering, Toronto, Ontario (Canada)

    1997-07-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  12. 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.

  13. 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

  14. Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil, E-mail: hgkim@kaeri.re.kr; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun

    2015-10-15

    A 3D laser coating technology using Cr powder was developed for Zr-based alloys considering parameters such as: the laser beam power, inert gas flow, cooling of Zr-based alloys, and Cr powder control. This technology was then applied to Zr cladding tube samples to study the effect of Cr coating on the high-temperature oxidation of Zr-based alloys in a steam environment of 1200 °C for 2000s. It was revealed that the oxide layer thickness formed on the Cr-coated tube surface was about 25-times lower than that formed on a Zircaloy-4 tube surface. In addition, both the ring compression and the tensile tests were performed to evaluate the adhesion properties of the Cr-coated sample. Although some cracks were formed on the Cr-coated layer, the Cr-coated layer had not peeled off after the two tests.

  15. High temperature mechanisms and kinetics of SiC oxidation under low partial pressures of oxygen: application to the fuel cladding of gas fast reactors

    International Nuclear Information System (INIS)

    Hun, N.

    2011-01-01

    Gas Fast Reactor (GFR) is one of the different Generation IV concepts under investigation for energy production. SiC/SiC composites are candidates of primary interest for a GFR fuel cladding use, thanks to good corrosion resistance among other properties. The mechanisms and kinetics of SiC oxidation under operating conditions have to be identified and quantified as the corrosion can decrease the mechanical properties of the composite. An experimental device has been developed to study the oxidation of silicon carbide under high temperature and low oxygen partial pressure. The results pointed out that not only parabolic oxidation, but also interfacial reactions and volatilization occur under such conditions. After determining the kinetics of each mechanism, as functions of oxygen partial pressure and temperature, the data are used for the modeling of the composites oxidation. The model will be used to predict the lifetime of the composite in operating conditions. (author) [fr

  16. Fabrication, properties, and microstructures of high Tc tapes and coils made from Ag-clad Bi-2223 superconductors

    International Nuclear Information System (INIS)

    Balachandran, U.; Iyer, A.N.; Youngdahl, C.A.; Motowidlo, L.R.; Hoehn, J.G.; Haldar, P.

    1993-07-01

    Bi-2223 precursor powders were prepared via a solid-state reaction using carbonates and oxides of Bi, Pb, Sr, Ca, and Cu. Results indicate that an in-situ reaction between constituent phases with tho formation of a transient liquid that is consumed during final heat treatment, is essential to obtain increased density with greater connectivity between the 2223 grains. Relative amounts of the constituent phases were adjusted in the powder by varying the calcination conditions, and the powder was then used to fabricate Ag-clad tapes by a powder-in-tube technique. By improving process conditions, transport critical current density (J c ) values greater than 4 x 10 4 A/cm 2 at 77 K and 2 x 10 5 A/cm 2 at 4.2 and 27 K have been obtained in short tape samples. Long tapes were cut into lengths upto 10 m long and used in parallel to fabricate small superconducting pancake coils. The coils were characterized at 77, 27 and 4.2 K and results are discussed

  17. Deformation behavior of laser welds in high temperature oxidation resistant Fe–Cr–Al alloys for fuel cladding applications

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G., E-mail: fieldkg@ornl.gov; Gussev, Maxim N., E-mail: gussevmn@ornl.gov; Yamamoto, Yukinori, E-mail: yamamotoy@ornl.gov; Snead, Lance L., E-mail: sneadll@ornl.gov

    2014-11-15

    Ferritic-structured Fe–Cr–Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe–(13–17.5)Cr–(3–4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  18. Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

    Science.gov (United States)

    Field, Kevin G.; Gussev, Maxim N.; Yamamoto, Yukinori; Snead, Lance L.

    2014-11-01

    Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  19. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  20. Electra-Clad

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-05-04

    The study relates to the use of building-integrated photovoltaics. The Electra-Clad project sought to use steel-based cladding as a substrate for direct fabrication of a fully integrated solar panel of a design similar to the ICP standard glass-based panel. The five interrelated phases of the project are described. The study successfully demonstrated that the principles of the panel design are achievable and sound. But, despite intensive trials, a commercially realistic solar performance has not been achieved: the main failing was the poor solar conversion efficiency as the active area of the panel was increased in size. The problem lies with the coating used on the steel cladding substrates and it was concluded that a new type of coating will be required. ICP Solar Technologies UK carried out the work under contract to the DTI.

  1. Nuclear fuel cladding material

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo.

    1982-01-01

    Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

  2. 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.

  3. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Jen, C.-K.; Ihara, I.

    1999-01-01

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  4. Microstructures and properties of low-chromium high corrosion-resistant TiC-VC reinforced Fe-based laser cladding layer

    International Nuclear Information System (INIS)

    Zhang, Hui; Zou, Yong; Zou, Zengda; Wu, Dongting

    2015-01-01

    Highlights: • The cladding layer with 3.0%Cr and 0.25%CeO 2 showed a good corrosion resistance. • Passive film formed on the cladding layer without Cr and CeO 2 was Fe 3 O 4 . • Fe 3 O 4 displayed p type semiconductivity. • Passive film formed on the cladding layer with Cr and CeO 2 was Fe(OH) 3 and Cr(OH) 3 . • Fe(OH) 3 displayed n type while Cr(OH) 3 displayed p type semiconductivity. - Abstract: Effects of 3.0 wt.%Cr and/or 0.25 wt.%CeO 2 on microstructures and properties of TiC-VC reinforced Fe-based cladding layer were investigated by using X-ray diffractometry (XRD), scanning electron microscopy (SEM), and electrochemical impedance spectroscopy (EIS). Passive films formed on cladding layers surface were investigated by using X-ray photoelectron spectroscopy (XPS) and Mott-Schottky analysis. Results showed that phases of cladding layers were α-Fe, γ-Fe, TiC, VC and TiVC 2 . There were no obvious effects of adding 3.0 wt.%Cr and/or 0.25 wt.%CeO 2 on cladding layers phases. The microstructure of the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO 2 was lath martensite and retained austenite. Microhardness of the cladding layer with 0.25 wt.%CeO 2 decreased slightly. Microhardness and corrosion resistance of the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO 2 both increased, the corrosion resistance increased 7.33 times while the EIS Nyquist spectrum transformed into a capacitive arc. The passive film formed on the cladding layer without Cr and CeO 2 was Fe 3 O 4 which displayed p type semiconductivity. The passive film formed on the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO 2 was composed of Fe(OH) 3 and Cr(OH) 3 , which displayed n and p type semiconductivity respectively

  5. Microstructures and properties of low-chromium high corrosion-resistant TiC-VC reinforced Fe-based laser cladding layer

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hui; Zou, Yong, E-mail: yzou@sdu.edu.cn; Zou, Zengda; Wu, Dongting

    2015-02-15

    Highlights: • The cladding layer with 3.0%Cr and 0.25%CeO{sub 2} showed a good corrosion resistance. • Passive film formed on the cladding layer without Cr and CeO{sub 2} was Fe{sub 3}O{sub 4}. • Fe{sub 3}O{sub 4} displayed p type semiconductivity. • Passive film formed on the cladding layer with Cr and CeO{sub 2} was Fe(OH){sub 3} and Cr(OH){sub 3}. • Fe(OH){sub 3} displayed n type while Cr(OH){sub 3} displayed p type semiconductivity. - Abstract: Effects of 3.0 wt.%Cr and/or 0.25 wt.%CeO{sub 2} on microstructures and properties of TiC-VC reinforced Fe-based cladding layer were investigated by using X-ray diffractometry (XRD), scanning electron microscopy (SEM), and electrochemical impedance spectroscopy (EIS). Passive films formed on cladding layers surface were investigated by using X-ray photoelectron spectroscopy (XPS) and Mott-Schottky analysis. Results showed that phases of cladding layers were α-Fe, γ-Fe, TiC, VC and TiVC{sub 2}. There were no obvious effects of adding 3.0 wt.%Cr and/or 0.25 wt.%CeO{sub 2} on cladding layers phases. The microstructure of the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO{sub 2} was lath martensite and retained austenite. Microhardness of the cladding layer with 0.25 wt.%CeO{sub 2} decreased slightly. Microhardness and corrosion resistance of the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO{sub 2} both increased, the corrosion resistance increased 7.33 times while the EIS Nyquist spectrum transformed into a capacitive arc. The passive film formed on the cladding layer without Cr and CeO{sub 2} was Fe{sub 3}O{sub 4} which displayed p type semiconductivity. The passive film formed on the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO{sub 2} was composed of Fe(OH){sub 3} and Cr(OH){sub 3}, which displayed n and p type semiconductivity respectively.

  6. Influence Of The Laser Cladding Strategies On The Mechanical Properties Of Inconel 718

    International Nuclear Information System (INIS)

    Lamikiz, A.; Tabernero, I.; Ukar, E.; Lopez de Lacalle, L. N.; Delgado, J.

    2011-01-01

    This work presents different experimental results of the mechanical properties of Inconel registered 718 test parts built-up by laser cladding. Recently, turbine manufacturers for aeronautical sector have presented high interest on laser cladding processes. This process allows building fully functional structures on superalloys, such as Inconel registered 718, with high flexibility on complex shapes. However, there is limited data on mechanical properties of the laser cladding structures. Moreover, the available data do not include the influence of process parameters and laser cladding strategies. Therefore, a complete study of the influence of the laser cladding parameters and mainly, the variation of the tensile strength with the laser cladding strategy is presented. The results show that there is a high directionality of mechanical properties, depending on the strategies of laser cladding process. In other words, the test parts show a fiber -like structure that should be considered on the laser cladding strategy selection.

  7. 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)

  8. Elliptical As2Se3 filled core ultra-high-nonlinearity and polarization-maintaining photonic crystal fiber with double hexagonal lattice cladding

    Science.gov (United States)

    Li, Feng; He, Menghui; Zhang, Xuedian; Chang, Min; Wu, Zhizheng; Liu, Zheng; Chen, Hua

    2018-05-01

    A high birefringence and ultra-high nonlinearity photonic crystal fiber (PCF) is proposed, which is composed of an elliptical As2Se3-doped core and an inner cladding with hexagonal lattice. Optical properties of the PCF are simulated by the full-vector finite element method. The simulation results show that the high birefringence of ∼0.33, ultra-high-nonlinearity coefficient of 300757 W-1km-1 and the low confinement loss can be achieved in the proposed PCF simultaneously at the wavelength of 1.55 μm. Furthermore, by comparison with the other two materials (80PbO•20Ga2O3, As2S3) filled in the core, the As2Se3-doped PCF is found to have the highest birefringence and nonlinearity due to its higher refractive index and nonlinear refractive index. The flattened dispersion feature, as well as the low confinement loss of the proposed PCF structure make it suitable as a wide range of applications, such as the coherent optical communications, polarization-maintaining and nonlinear optics, etc.

  9. Corrosion behavior of duplex and reference cladding in NPP Grohnde

    International Nuclear Information System (INIS)

    Besch, O.A.; Yagnik, S.K.; Eucken, C.M.; Bradley, E.R.

    1996-01-01

    The Nuclear Fuel Industry Research (NFIR) Group undertook a lead test assembly (LTA) program in NPP Grohnde PWR in Germany to assess the corrosion performance of duplex and reference cladding. Two identical 16 by 16 LTAs, each containing 32 peripheral test rods, completed four reactor cycles, reaching a peak rod burnup of 46 MWd/kgU. The results from poolside examinations performed at the end of each cycle, together with power histories and coolant chemistry, are reported. Five different cladding materials were characterized during fabrication. The corrosion performance of the cladding materials was tracked in long-term tests in high-pressure, high-temperature autoclaves. The relative ranking of corrosion behavior in such tests corresponded well with the in-reactor corrosion performance. The extent and distribution of hydriding in duplex and reference specimens during the autoclave testing has been characterized. The in-reactor corrosion data indicate that the low-tin Zircaloy-4 reference cladding, R2, had an improved corrosion resistance compared to high-tin Zircaloy-4 reference cladding, R1. Two types of duplex cladding, D1 (Zr-2.5% Nb) and D2 (Zr-0.4% Fe-0.5% Sn), showed an even further improvement in corrosion resistance compared to R2 cladding. The third duplex cladding, D3 (Zr-4 + 1.0% Nb), had significantly less corrosion resistance, which was inferior to R1. The in-reactor and out-reactor corrosion performances have been ranked

  10. 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.)

  11. 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)

  12. 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)

  13. 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.

  14. 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

  15. 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).

  16. 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.

  17. Study on process of laser cladded nuclear valve parts

    International Nuclear Information System (INIS)

    Zhang Chunliang

    2000-01-01

    The microstructure and performances of the Co-base alloy coatings that are formed by laser cladding, plasma spurt welding and arc surfacing on the nuclear valve-sealing surface have been studied and compared. The combination costs of laser cladding, plasma spurt welding and arc, surfacing have been analyzed and compared. The results showed that the laser cladding processing has the advantages of high efficiency, low energy cost, a little machining allowance, high rate of finished products and low combination cost, compared with plasma spurt welding processing and arc surfacing processing. The laser cladding technology can improve the qualities of nuclear valve parts and increase their service life. Therefore, the laser cladding processing is a new technology with developing potential

  18. 2 µm high-power dissipative soliton resonance in a compact σ-shaped Tm-doped double-clad fiber laser

    Science.gov (United States)

    Du, Tuanjie; Li, Weiwei; Ruan, Qiujun; Wang, Kaijie; Chen, Nan; Luo, Zhengqian

    2018-05-01

    We report direct generation of a high-power, large-energy dissipative soliton resonance (DSR) in a 2 µm Tm-doped double-clad fiber laser. A compact σ-shaped cavity is formed by a fiber Bragg grating and a 10/90 fiber loop mirror (FLM). The 10/90 FLM is not only used as an output mirror, but also acts as a nonlinear optical loop mirror for initiating mode locking. The mode-locked laser can deliver high-power, nanosecond DSR pulses at 2005.9 nm. We further perform a comparison study of the effect of the FLM’s loop length on the mode-locking threshold, peak power, pulse energy, and optical spectrum of the DSR pulses. We achieve a maximum average output power as high as 1.4 W, a maximum pulse energy of 353 nJ, and a maximum peak power of 84 W. This is, to the best of our knowledge, the highest power for 2 µm DSR pulses obtained in a mode-locked fiber laser.

  19. Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

    Science.gov (United States)

    Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

    2017-12-01

    High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

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

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  1. Inpile (in PWR) testing of cladding materials

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    As an introduction, the reasons to perform inpile tests are depicted. An overview over general inpile test procedure is given, and test details which are necessary for the development of new alloys for high burnup claddings, like sample geometries and measuring techniques for inpile corrosion testing, are described in detail. Tests for the creep and length change behavior of cladding tubes are described briefly. Finally, conclusions are drawn and literature citations for further test details are given. (author). 9 refs., 2 tabs., 17 figs

  2. 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

  3. 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.

  4. Application of Coating Technology for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

  5. Corrosion behaviour of Zircaloy 4 fuel cans for high burnup in EdF PWRs

    International Nuclear Information System (INIS)

    Blat, M.; Kerrec, O.; Bourgoin, J.; Vrignaud, E.; Amanrich, H.

    1994-01-01

    Uniform corrosion of fuel cladding could be a limitation for burn-up enhancement. First, the oxide thickness measured on fuel cladding for high burn-up has been compared to the prediction of the EDF code, CYRANO 2E. A comparative metallurgical characterization has been also performed on samples which were oxidized in pile and in autoclave. Then, laboratories studies have been launched for a better understanding of the corrosion mechanisms. A reflection was proposed on the two main theoretical concepts proposed for these mechanisms. Their kinetics could be controlled by transfers in liquid medium (electrolyte) or in solid medium (compact oxide). For the first topic, a nanoscopic characterization of the oxide is in progress, using Atomic Force Microscope. The first results are presented. In the second case, an electrochemical approach (impedance spectroscopy and voltametry) is developed in our laboratories. The obtained results could give some new keys in order to understand the influence of some parameters (alloys composition, coolant chemistry,...). (authors). 7 figs., 1 tab., 7 refs

  6. Effects of irradiation on initiation and crack-arrest toughness of two high-copper welds and on stainless steel cladding

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Iskander, S.K.; Haggag, F.M.

    1990-01-01

    The objective of the study on the high-copper welds is to determine the effect of neutron irradiation on the shift and shape of the ASME K Ic and K Ia toughness curves. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Compact specimens fabricated from these welds were irradiated at a nominal temperature of 288 degree C to fluences from 1.5 to 1.9 x 10 19 neutrons/cm 2 (>1 MeV). The fracture toughness test results show that the irradiation-induced shifts at 100 MPa/m were greater than the Charpy 41-J shifts by about 11 and 18 degree C. Mean curve fits indicate mixed results regarding curve shape changes, but curves constructed as lower boundaries to the data do indicate curves of lower slopes. A preliminary evaluation of the crack-arrest results shows that the neutron-irradiation induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower bound curves (for the range of test temperatures covered), compared to those of the ASME K Ia curve did not appear to have been altered by the irradiation. Three-wire stainless steel weld overlay cladding was irradiated at 288 degree C to fluences of 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV). Charpy 41-J temperature shifts of 13 and 28 degree C were observed, respectively. For the lower fluence only, 12.7-mm thick compact specimens showed decreases in both J Ic and the tearing modulus. Comparison of the fracture toughness results with typical plate and a low upper-shelf weld reveals that the irradiated stainless steel cladding possesses low ductile initiation fracture toughness comparable to the low upper-shelf weld. 8 refs., 12 figs., 2 tabs

  7. Reduction of Bragg-grating-induced coupling to cladding modes

    DEFF Research Database (Denmark)

    Berendt, Martin Ole; Bjarklev, Anders Overgaard; Soccolich, C.E.

    1999-01-01

    gratings in a depressed-cladding fiber are compared with simulations. The model gives good agreement with the measured transmission spectrum and accounts for the pronounced coupling to asymmetrical cladding modes, even when the grating is written with the smallest possible blaze. The asymmetry causing...... this is accounted for by the unavoidable attenuation of the UV light. It is found for the considered fiber designs that a high numerical-aperture fiber increases the spectral separation between the Bragg resonance and the onset of cladding-mode losses. A depressed-cladding fiber reduces the coupling strength......We discuss fiber designs that have been suggested for the reduction of Bragg-grating induced coupling to cladding modes. The discussion is based on a theoretical approach that includes the effect of asymmetry in the UV-induced index grating, made by UV-side writing. Experimental results from...

  8. Polarization characteristics of double-clad elliptical fibers.

    Science.gov (United States)

    Zhang, F; Lit, J W

    1990-12-20

    A scalar variational analysis based on a Gaussian approximation of the fundamental mode of a double-clad elliptical fiber with a depressed inner cladding is studied. The polarization properties and graphic results are presented; they are given in terms of three parameters: the ratio of the major axis to the minor axis of the core, the ratio of the inner cladding major axis to the core major axis, and the difference between the core index and the inner cladding index. The variations of both the spot size and the field intensity with core ellipticity are examined. It is shown that high birefringence and dispersion-free orthogonal polarization modes can be obtained within the single-mode region and that the field intensity distribution may be more confined to the fiber center than in a single-clad elliptical fiber.

  9. Microstructure and high-temperature oxidation resistance of TiN/Ti3Al intermetallic matrix composite coatings on Ti6Al4V alloy surface by laser cladding

    Science.gov (United States)

    Zhang, Xiaowei; Liu, Hongxi; Wang, Chuanqi; Zeng, Weihua; Jiang, Yehua

    2010-11-01

    A high-temperature oxidation resistant TiN embedded in Ti3Al intermetallic matrix composite coating was fabricated on titanium alloy Ti6Al4V surface by 6kW transverse-flow CO2 laser apparatus. The composition, morphology and microstructure of the laser clad TiN/Ti3Al intermetallic matrix composite coating were characterized by optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD) and energy dispersive spectrometer (EDS). In order to evaluate the high-temperature oxidation resistance of the composite coatings and the titanium alloy substrate, isothermal oxidation test was performed in a conventional high-temperature resistance furnace at 600°C and 800°C respectively. The result shows that the laser clad intermetallic composite coating has a rapidly solidified fine microstructure consisting of TiN primary phase (granular-like, flake-like, and dendrites), and uniformly distributed in the Ti3Al matrix. It indicates that a physical and chemical reaction between the Ti powder and AlN powder occurred completely under the laser irradiation. In addition, the microhardness of the TiN/Ti3Al intermetallic matrix composite coating is 844HV0.2, 3.4 times higher than that of the titanium alloy substrate. The high-temperature oxidation resistance test reveals that TiN/Ti3Al intermetallic matrix composite coating results in the better modification of high-temperature oxidation behavior than the titanium substrate. The excellent high-temperature oxidation resistance of the laser cladding layer is attributed to the formation of the reinforced phase TiN and Al2O3, TiO2 hybrid oxide. Therefore, the laser cladding TiN/Ti3Al intermetallic matrix composite coating is anticipated to be a promising oxidation resistance surface modification technique for Ti6Al4V alloy.

  10. Analysis of atomic distribution in as-fabricated Zircaloy-2 claddings by atom probe tomography under high-energy pulsed laser

    Energy Technology Data Exchange (ETDEWEB)

    Sawabe, T., E-mail: sawabe@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Sonoda, T.; Kitajima, S. [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Kameyama, T. [Tokai University, Department of Nuclear Engineering, Kitakaname 4-1-1, Hiratsuka, Kanagawa 259-1292 (Japan)

    2013-11-15

    The properties of second-phase particles (SPPs) in Zircaloy-2 claddings are key factors influencing the corrosion resistance of the alloy. The chemical compositions of Zr (Fe, Cr){sub 2} and Zr{sub 2}(Fe, Ni) SPPs were investigated by means of pulsed laser atom probe tomography. In order to prevent specimen fracture and to analyse wide regions of the specimen, the pulsed laser energy was increased to 2.0 nJ. This gave a high yield of average of 3 × 10{sup 7} ions per specimen. The Zr (Fe, Cr){sub 2} SPPs contained small amounts of Ni and Si atoms, while in Zr{sub 2}(Fe, Ni) SPPs almost all the Si was concentrated and the ratio of Zr: (Fe + Ni + Si) was 2:1. Atomic concentrations of the Zr-matrix and the SPPs were identified by two approaches: the first by using all the visible peaks of the mass spectrum and the second using the representative peaks with the natural abundance of the corresponding atoms. It was found that the change in the concentration between the Zr-matrix and the SPPs can be estimated more accurately by the second method, although Sn concentration in the Zr{sub 2}(Fe, Ni) SPPs is slightly overestimated.

  11. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  12. Assessment of oxygen diffusion coefficients by studying high-temperature oxidation behaviour of Zr1Nb fuel cladding in the temperature range of 1100–1300 °C

    Energy Technology Data Exchange (ETDEWEB)

    Négyesi, M., E-mail: negy@seznam.cz [Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Trojanova 13, 120 00 Praha 2 (Czech Republic); UJP PRAHA a.s., Nad Kamínkou 1345, 156 10 Praha – Zbraslav (Czech Republic); Chmela, T. [UJP PRAHA a.s., Nad Kamínkou 1345, 156 10 Praha – Zbraslav (Czech Republic); Veselský, T. [Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Trojanova 13, 120 00 Praha 2 (Czech Republic); Krejčí, J. [Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Trojanova 13, 120 00 Praha 2 (Czech Republic); CHEMCOMEX Praha a.s., Elišky Přemyslovny 379, 156 10 Praha – Zbraslav (Czech Republic); Novotný, L.; Přibyl, A. [UJP PRAHA a.s., Nad Kamínkou 1345, 156 10 Praha – Zbraslav (Czech Republic); Bláhová, O. [New Technologies Research Centre, University of West Bohemia, Univerzitní 8, 306 14 Plzeň (Czech Republic); Burda, J. [NRI Rez plc, Husinec-Řež 130, 250 68 Řež (Czech Republic); Siegl, J. [Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Trojanova 13, 120 00 Praha 2 (Czech Republic); Vrtílková, V. [UJP PRAHA a.s., Nad Kamínkou 1345, 156 10 Praha – Zbraslav (Czech Republic)

    2015-01-15

    The paper deals with high-temperature steam oxidation behaviour of Zr1Nb fuel cladding. First of all, comprehensive experimental program was conducted to provide sufficient experimental data, such as the thicknesses of evolved phase layers and the overall weight gain kinetics, as well as the oxygen concentration and nanohardness values at phase boundaries. Afterwards, oxygen diffusion coefficients in the oxide, in the α-Zr(O) layer, in the double-phase (α + β)-Zr region, and in the β-phase region have been estimated based on the experimental data employing analytical solution of the multiphase moving boundary problem, assuming the equilibrium conditions being fulfilled at the interface boundaries. Eventually, the determined oxygen diffusion coefficients served as input into the in-house numerical code, which was designed to predict the high-temperature oxidation behaviour of Zr1Nb fuel cladding. Very good agreement has been achieved between the numerical calculations and the experimental data.

  13. Highly Sensitive Refractive Index Sensor Based on a Cladding-Etched Thin-Core Fiber Sandwiched between Two Single-Mode Fibers

    International Nuclear Information System (INIS)

    Xu Ben; Li Yi; Dong Xin-Yong; Jin Shang-Zhong; Zhang Zai-Xuan

    2012-01-01

    A refractive index (RI) sensor based on a cladding-etched thin-core single-mode fiber (TCSMF) sandwiched between two single-mode fibers is demonstrated. The experimental results show that the sensitivity, within the RI range of 1.333–1.340, is enhanced at least 6 times by etching. It increases with the surrounding RI and reaches 857.5 nm/RIU at RI of 1.3684, and it can be expected to be higher with the decrease of the cladding diameter of the TCSMF

  14. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...... for various biosensing applications, however a thorough study of the sensor configurations has not been presented, but is the main subject of this thesis. Optical sensors are generally well suited for bio-sensing asthey show high sensitivity and give an immediate response for minute changes in the refractive...... index of a sample, due to the high sensitivity of optical bio-sensors detection of non-labeled biological objects can be performed. The majority of opticalsensors presented in the literature and commercially available optical sensors are based on evanescent wave sensing, however most of these sensors...

  15. Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

    International Nuclear Information System (INIS)

    Luo Bangqi

    2012-01-01

    The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

  16. Development and application of preventive maintenance technique for pipes using laser cladding method

    International Nuclear Information System (INIS)

    Hatakenaka, Hiroaki; Yamadera, Masao; Shiraiwa, Takanori.

    1995-01-01

    A laser cladding method which produces a highly corrosion-resisting coating (cladding) on the surface of the material was developed for the purpose of preventing stress corrosion cracking (SCC) in the austenitic stainless steel (Type 304). In this method, metallic powder paste is applied on the inner surface of pipes, and then a YAG laser beam is irradiated to the paste, which melts and forms a clad with excellent corrosion resistance. Recently, the laser cladding method was practically and successfully applied to the actual nuclear power plant in Japan. This report describes this laser cladding technique, the equipment, and actual works in the field. (author)

  17. Slurry Erosion Behavior of AlxCoCrFeNiTi0.5 High-Entropy Alloy Coatings Fabricated by Laser Cladding

    Directory of Open Access Journals (Sweden)

    Jianhua Zhao

    2018-02-01

    Full Text Available High-entropy alloys (HEAs have gained extensive attention due to their excellent properties and the related scientific value in the last decade. In this work, AlxCoCrFeNiTi0.5 HEA coatings (x: molar ratio, x = 1.0, 1.5, 2.0, and 2.5 were fabricated on Q345 steel substrate by laser-cladding process to develop a practical protection technology for fluid machines. The effect of Al content on their phase evolution, microstructure, and slurry erosion performance of the HEA coatings was studied. The AlxCoCrFeNiTi0.5 HEA coatings are composed of simple face-centered cubic (FCC, body-centered cubic (BCC and their mixture phase. Slurry erosion tests were conducted on the HEA coatings with a constant velocity of 10.08 m/s and 16–40 meshs and particles at impingement angles of 15, 30, 45, 60 and 90 degrees. The effect of three parameters, namely impingement angle, sand concentration and erosion time, on the slurry erosion behavior of AlxCoCrFeNiTi0.5 HEA coatings was investigated. Experimental results show AlCoCrFeNiTi0.5 HEA coating follows a ductile erosion mode and a mixed mode (neither ductile nor brittle for Al1.5CoCrFeNiTi0.5 HEA coating, while Al2.0CoCrFeNiTi0.5 and Al2.5CoCrFeNiTi0.5 HEA coatings mainly exhibit brittle erosion mode. AlCoCrFeNiTi0.5 HEA coating has good erosion resistance at all investigated impingement angles due to its high hardness, good plasticity, and low stacking fault energy (SFE.

  18. ;Study of secondary hydriding at high temperature in zirconium based nuclear fuel cladding tubes by coupling information from neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and laser induced breakdown spectroscopy microprobe

    Science.gov (United States)

    Brachet, Jean-Christophe; Hamon, Didier; Le Saux, Matthieu; Vandenberghe, Valérie; Toffolon-Masclet, Caroline; Rouesne, Elodie; Urvoy, Stéphane; Béchade, Jean-Luc; Raepsaet, Caroline; Lacour, Jean-Luc; Bayon, Guy; Ott, Frédéric

    2017-05-01

    This paper gives an overview of a multi-scale experimental study of the secondary hydriding phenomena that can occur in nuclear fuel cladding materials exposed to steam at high temperature (HT) after having burst (loss-of-coolant accident conditions). By coupling information from several facilities, including neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and micro laser induced breakdown spectroscopy, it was possible to map quantitatively, at different scales, the distribution of oxygen and hydrogen within M5™ clad segments having experienced ballooning and burst at HT followed by steam oxidation at 1100 and 1200 °C and final direct water quenching down to room temperature. The results were very reproducible and it was confirmed that internal oxidation and secondary hydriding at HT of a cladding after burst can lead to strong axial and azimuthal gradients of hydrogen and oxygen concentrations, reaching 3000-4000 wt ppm and 1.0-1.2 wt% respectively within the β phase layer for the investigated conditions. Consistent with thermodynamic and kinetics considerations, oxygen diffusion into the prior-β layer was enhanced in the regions highly enriched in hydrogen, where the α(O) phase layer is thinner and the prior-β layer thicker. Finally the induced post-quenching hardening of the prior-β layer was mainly related to the local oxygen enrichment. Hardening directly induced by hydrogen was much less significant.

  19. Study and Behaviour of Prefabricated Composite Cladding

    Science.gov (United States)

    Sai Avinash, P.; Thiagarajan, N.; Santhi, A. S.

    2017-07-01

    The incessant population rise entailed for an expeditious construction at competitive prices that steered the customary path to the light weight structural components. This lead to construction of structural components using ferrocement. The load bearing structural cladding, sizing 3200x900x100 mm, is chosen for the study, which, is analyzed using the software ABAQUS 6.14 in accordance with the IS:875-87 Part1, IS:875-87 Part2, ACI 549R-97, ACI 318R-08 and NZS:3101-06 Part1 standards. The Ferrocement claddings (FCs) are fabricated to a scaled dimension of 400x115x38 mm. The light weight-high strength phenomena are corroborated by incorporating Glass Fibre Reinforced Polymer Laminates (GFRPL) of thickness 6mm, engineered with the aid of hand layup (wet layup) technique wielding epoxy resin, followed by curing under room temperature. The epoxy resin is employed for fastening ferrocement cladding with the Glass fiber reinforced polymer laminate, with the contemporary methodology. The compressive load carrying capacity of the amalgamated assembly, both in presence and absence of Glass Fibre Reinforced polymer laminates (GFRPL) on either side of Ferrocement cladding, has been experimented.

  20. Contribution to the study of the pseudobinary Zr1Nb-Oxygen phase diagram by local oxygen measurements of Zr1Nb fuel cladding after high temperature oxidation

    Czech Academy of Sciences Publication Activity Database

    Negyesi, M.; Burda, J.; Klouček, V.; Lorinčík, Jan; Sopoušek, J.; Kabátová, J.; Novotný, L.; Linhart, S.; Chmela, T.; Siegl, J.; Vrtílková, V.

    2012-01-01

    Roč. 420, 1-3 (2012), s. 314-319 ISSN 0022-3115 Institutional research plan: CEZ:AV0Z20670512 Keywords : Zr1Nb * oxygen * fuel cladding Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 1.211, year: 2012

  1. Monitoring layer-by-layer assembly of polyelectrolyte multi-layers using high-order cladding mode in long-period fiber gratings

    Czech Academy of Sciences Publication Activity Database

    Tian, F.; Kaňka, Jiří; Li, X.; Du, H.

    -, č. 196 (2014), s. 475-479 ISSN 0925-4005 R&D Projects: GA MŠk(CZ) LH11038 Institutional support: RVO:67985882 Keywords : Layer-by-layer assembly * Polyelectrolyte * Cladding mode Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 4.097, year: 2014

  2. Siemens advance PWR fuel assemblies (HTP) and cladding

    International Nuclear Information System (INIS)

    Stout, R. B.; Woods, K. N.

    1997-01-01

    This paper describes the key features of the Siemens HTP (High Thermal Performance) fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available

  3. Development and characterization of laser clad high temperature self-lubricating wear resistant composite coatings on Ti–6Al–4V alloy

    International Nuclear Information System (INIS)

    Liu, Xiu-Bo; Meng, Xiang-Jun; Liu, Hai-Qing; Shi, Gao-Lian; Wu, Shao-Hua; Sun, Cheng-Feng; Wang, Ming-Di; Qi, Long-Hao

    2014-01-01

    Highlights: • A novel high temperature self-lubricating anti-wear composite coating was fabricated. • Reinforced carbides as well as self-lubricating sulfides were in situ synthesized. • Microhardness of the Ti–6Al–4V substrate was significantly improved. • Friction coefficient and wear rate of the composite coating were greatly reduced. - Abstract: To enhance the wear resistance and friction-reducing capability of titanium alloy, a process of laser cladding γ-NiCrAlTi/TiC + TiWC 2 /CrS + Ti 2 CS coatings on Ti–6Al–4V alloy substrate with preplaced NiCr/Cr 3 C 2 –WS 2 mixed powders was studied. A novel coating without cracks and few pores was obtained in a proper laser processing. The composition and microstructure of the fabricated coating were examined by X-ray diffraction (XRD), scanning electron microscope (SEM), energy dispersive spectroscopy (EDS) techniques, and tribological properties were evaluated using a ball-on-disc tribometer under dry sliding wear test conditions at 20 °C (room-temperature), 300 °C, 600 °C, respectively. The results show that the coating has unique microstructure consisting of α-Ti, TiC, TiWC 2 , γ-NiCrAlTi, Ti 2 CS and CrS phases. Average microhardness of the composite coating is 1005 HV 0.2 , which is about 3-factor higher than that of Ti–6Al–4V substrate (360 HV 0.2 ). The friction coefficient and wear rate of the coating are greatly decreased due to the combined effects of the dominating anti-wear capabilities of reinforced TiC and TiWC 2 carbides and the CrS and Ti 2 CS sulfides which have excellent self-lubricating property

  4. The Phase Evolution and Property of FeCoCrNiAlTix High-Entropy Alloying Coatings on Q253 via Laser Cladding

    Directory of Open Access Journals (Sweden)

    Bin He

    2017-09-01

    Full Text Available High-entropy alloys (HEAs are emerging as a hot research frontier in the metallic materials field. The study on the effect of alloying elements on the structure and properties of HEAs may contribute to the progress of the research and accelerate the application in actual production. FeCoCrNiAlTix (x = 0, 0.25, 0.5, 0.75, and 1 in at.%, respectively HEA coatings with different Ti concentrations were produced on Q235 steel via laser cladding. The constituent phases, microstructure, hardness, and wear resistance of the coatings were investigated by XRD, SEM, microhardness tester and friction-wear tester, respectively. The results show that the structure of the coating is a eutectic microstructure of FCC and BCC1 at x = 0. The structure of coatings consists of both proeutectic FCC phase and the eutectic structure of BCC1 and BCC2. With the continuous addition of Ti, the amount of eutectic structure decreases. The average hardness of the FeCoCrNiAlTix HEA coatings at x = 0, 0.25, 0.5, 0.75, and 1 are 432.73 HV, 548.81 HV, 651.03 HV, 769.20 HV, and 966.29 HV, respectively. The hardness of coatings increases with the addition of Ti, where the maximum hardness is achieved for the HEA at x = 1. The wear resistance of the HEA coatings is enhanced with the addition of Ti, and the main worn mechanism is abrasive wear.

  5. Estimation of penetration depth of fission products in cladding Hull

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

    2005-01-01

    A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

  6. CASTI handbook of cladding technology. 2. ed.

    International Nuclear Information System (INIS)

    Smith, L.; Celant, M.

    2000-01-01

    This updated (2000) CASTI handbook covers all aspects of clad products - the different means of manufacture, properties and applications in various industries. Topics include: an introduction to cladding technology, clad plate, clad pipes, bends, clad fittings, specification requirements of clad products, welding clad products, clad product application and case histories from around the world. Unique to this book is the documentation of case histories of major cladding projects from around the world and how the technology of that day has withstood the demands of time. Filled with over 100 photos and graphics illustrating the various cladding technology examples and products, this book truly documents the most recent technologies in the field of cladding technology used worldwide

  7. Investigations on dry sliding of laser cladded aluminum bronze

    Directory of Open Access Journals (Sweden)

    Freiße Hannes

    2016-01-01

    Full Text Available The aim of this study was to investigate the tribological behaviour of laser cladded aluminum bronze tool surfaces for dry metal forming. In a first part of this work a process window for cladding aluminum bronze on steel substrate was investigated to ensure a low dilution. Therefore, the cladding speed, the powder feed rate, the laser power and the distance between the process head and the substrate were varied. The target of the second part was to investigate the influence of different process parameters on the tribological behaviour of the cladded tracks. The laser claddings were carried out on both aluminum bronze and cold work tool steel as substrate materials. Two different particle sizes of the cladding powder material were used. The cladding speed was varied and a post-processing laser remelting treatment was applied. It is shown that the tribological behaviour of the surface in a dry oscillating ball-on-plate test is highly dependent on the substrate material. In the third part a deep drawing tool was additively manufactured by direct laser deposition. Furthermore, the tool was applied to form circular cups with and without lubrication.

  8. 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

  9. The effect of laser process parameters on microstructure and dilution rate of cladding coatings

    Science.gov (United States)

    Bin, Liu; Heping, Liu; Xingbin, Jing; Yuxin, Li; Peikang, Bai

    2018-02-01

    In order to broaden the range of application of Q235 steel, it is necessary to repair the surface of steel. High performance 316L stainless steel coating was successfully obtained on Q235 steel by laser cladding technology. The effect of laser cladding parameters on the geometrical size and appearance of single cladding layer was investigated. The experimental results show that laser current has an important influence on the surface morphology of single channel cladding. When the current is from 155A to 165A, the cladding coating becomes smooth. The laser current has an effect on the geometric cross section size and dilution rate of single cladding. The results revealed that with the rising of laser current, the width, height and depth of layer increase gradually. With the rising of laser current, the dilution rate of cladding layer is gradually increasing.

  10. Oxide thickness measurement technique for duplex-layer Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    McClelland, R.G.; O'Leary, P.M.

    1992-01-01

    Siemens Nuclear Power Corporation (SNP) is investigating the use of duplex-layer Zircaloy-4 tubing to improve the waterside corrosion resistance of cladding for high-burnup pressurized water reactor (PWR) fuel designs. Standard SNP PWR cladding is typically 0.762-mm (0.030-in.)-thick Zircaloy-4. The SNP duplex cladding is nominally 0.660-mm (0.026-in.)-thick Zircalloy-4 with an ∼0.102-mm (0.004-in.) outer layer of another, more corrosion-resistant, zirconium-based alloy. It is common industry practice to monitor the in-reactor corrosion behavior of Zircaloy cladding by using an eddy-current 'lift-off' technique to measure the oxide thickness on the outer surface of the fuel cladding. The test program evaluated three different cladding samples, all with the same outer diameter and wall thickness: Zircaloy-4 and duplex clad types D2 and D4

  11. High Temperature Mechanical Properties, Fractography and Synchrotron Studies of ATF clad materials from the UCSB-NSUF Irradiations.

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sprouster, David [Brookhaven National Lab. (BNL), Upton, NY (United States); Ecker, Lynne [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-02-23

    A variety of tensile samples of Ferritic and Oxide Dispersion Strengthened (ODS or nanostructured ferritic) steels were placed the ATR reactor over 2 years achieving doses of roughly 4-6 dpa at temperatures of roughly 290°C. Samples were shipped to Wing 9 in the CMR facility at Los Alamos National Laboratory and imaged then tested in tension. This report summarizes the room temperature tensile tests, the elevated temperature tensile tests (300°C) and fractography and reduction of area calculations on those samples. Additionally small samples were cut from the undeformed grip section of these tensile samples and sent to the NSLS synchrotron for high energy X-ray analysis, initial results will be described here.

  12. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-01-01

    The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

  13. 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.)

  14. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

    This volume presents new methodologies for the design of dimension stone based on the concepts of structural design while preserving the excellence of stonemasonry practice in façade engineering. Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements. Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  15. 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)

  16. Explosion Clad for Upstream Oil and Gas Equipment

    Science.gov (United States)

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

  17. Explosion Clad for Upstream Oil and Gas Equipment

    International Nuclear Information System (INIS)

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO 2 and/or H 2 S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

  18. Broadband generation by multiple four-wave mixing process due to ASE Q-switching in high-power double-clad ytterbium-doped fiber amplifier

    Science.gov (United States)

    Chowdhury, Sourav D.; Shekhar, Nishant; Saha, Maitreyee; Sen, Ranjan; Pal, Mrinmay

    2014-11-01

    Broadband output from 1060nm to 1700nm and cascaded four-wave mixing generated red light pulsing is observed in a fiber amplifier set up consisting of a 5.5m double clad, double D shaped Ytterbium doped fiber, a single clad passive fiber for excess pump absorption and a splitter, both with and without a CW seed. Self-pulsing occurs from ASE due to passive Q-switching by saturable absorption effect of the active fiber and also depends on splice loss. The pulses generate broadband output by multiple four-wave mixing process with maximum broadening efficiency near 1300nm which is the zero dispersion wavelength for silica fiber. Pulses traveling both in forward and backward direction have enough peak power and energy to damage splice points and fiber components. When seeded the self-pulsing and broadband generation is often suppressed but again generate at increased pump powers.

  19. Temperature-referenced high-sensitivity point-probe optical fiber chem-sensors based on cladding etched fiber Bragg gratings

    OpenAIRE

    Zhou, Kaiming; Chen, Xianfeng F.; Zhang, Lin; Bennion, Ian

    2004-01-01

    Point-probe optical fiber chem-sensors have been implemented using cladding etched fiber Bragg gratings. The sensors possess refractive index sensing capability that can be utilized to measure chemical concentrations. The Bragg wavelength shift reaches 8 nm when the index of surrounding medium changes from 1.33 to 1.44, giving maximum sensitivity more than 10 times higher than that of previously reported devices. More importantly, the dual-grating configuration of the point-probe sensors offe...

  20. High-temperature steam-oxidation behavior of Zr-1Nb-1Sn-0.1Fe cladding tube at temperatures of 800-1000

    International Nuclear Information System (INIS)

    Lee, Cheol Min; Cho, Tae Won; Jeong, Gwan Yoon; Kim, Mi Jin; Kim, Ji Hyeon; Lee, Hee Jae; Sohn, Dong Seong; Mok, Yong Kyoon

    2016-01-01

    To prevent cladding failure, NRC issued a regulation Title 10 § 50.46, which specifies cladding temperature of 1204 .deg. C and 17% ECR should not be exceeded. The fundamental reason of the mechanical degradation of cladding is the formation of the oxide which is brittle. Theoretically, the oxide layer is formed following parabolic rate. However, from many experiments, sub-parabolic rates are often observed. There have been many suggestions so far; chemical and stress gradient across the oxide layer could initiate the sub-parabolic rate, the phase transformation of Zirconium dioxide from tetragonal to monoclinic could be the reason, change of the grain size of Zirconium dioxide could cause the cubic oxidation rate, and there is a suggestion that if electron migration is the major mechanism of the oxide growth, then the subparabolic rate can show up. However, the reason why the sub-parabolic rate appears is still not certain. Another important degradation mechanism is breakaway oxidation. A clear explanation that why the breakaway oxidation appears is still not clear. Most of the people believe the phase transformation of Zirconium dioxide cause instability within the oxide, which causes breakaway oxidation to appear. However, how much effect is caused from the phase transformation is not so sure. In this study, detailed analysis about the oxidation kinetics and the breakaway oxidation of Zr-1Nb-1Sn- 0.1Fe were carried out at temperatures between 800 - 1000 .deg. C.

  1. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  2. Neutron-induced helium implantation in GCFR cladding

    International Nuclear Information System (INIS)

    Yamada, H.; Poeppel, R.B.; Sevy, R.H.

    1980-10-01

    The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10 10 He/cm 2 .s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 μm, more than 99% of helium particles are implanted in the first 2-μm-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding

  3. Composite polymer: Glass edge cladding for laser disks

    Science.gov (United States)

    Powell, H.T.; Wolfe, C.A.; Campbell, J.H.; Murray, J.E.; Riley, M.O.; Lyon, R.E.; Jessop, E.S.

    1987-11-02

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation. 18 figs.

  4. Composite polymer-glass edge cladding for laser disks

    Science.gov (United States)

    Powell, Howard T.; Riley, Michael O.; Wolfe, Charles R.; Lyon, Richard E.; Campbell, John H.; Jessop, Edward S.; Murray, James E.

    1989-01-01

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation.

  5. Comparison of fiber lasers based on distributed side-coupled cladding-pumped fibers and double-cladding fibers.

    Science.gov (United States)

    Huang, Zhihe; Cao, Jianqiu; Guo, Shaofeng; Chen, Jinbao; Xu, Xiaojun

    2014-04-01

    We compare both analytically and numerically the distributed side-coupled cladding-pumped (DSCCP) fiber lasers and double cladding fiber (DCF) lasers. We show that, through optimization of the coupling and absorbing coefficients, the optical-to-optical efficiency of DSCCP fiber lasers can be made as high as that of DCF lasers. At the same time, DSCCP fiber lasers are better than the DCF lasers in terms of thermal management.

  6. AGR fuel pin pellet-clad interaction failure limits and activity release fractions

    International Nuclear Information System (INIS)

    Hughes, H.; Hargreaves, R.

    1985-01-01

    The limiting conditions beyond which pellet-clad interaction can flail AGR fuel are described. They have been determined by many experiments involving post-irradiation examination and testing, loop experiments and cycling and up-rating of both individual fuel stringers and the whole WAGR core. The mechanisms causing this interaction are well understood and are quantitatively expressed in computer codes. Strain concentration effects over fuel cracks determine power cycling endurance while additional strain concentrations at clad ridges and from cross pin temperature gradients contribute to up-rating failures. An equation summarising tube burst test data so as to determine the ductility available at any transient is given. The hollow fuel and more ductile clad of the Civil AGR fuel pins leads to a much improved performance over the original fuel design. The Civil AGRs operate well within these limiting conditions and substantial increases beyond the design burn-up are confidently expected. The activity release on pin failure and its development during continued operation of failed fuel have also been investigated. A retention of radioiodine and caesium of 90-99% compared to the noble gases has been demonstrated. Measured fission gas releases into the free volume of Civil AGR fuel pins have been very low (< 0.1%)

  7. Tailoring nonlinearity and dispersion of photonic crystal fibers using hybrid cladding

    International Nuclear Information System (INIS)

    Zhao-lun, Liu; Lan-tian, Hou; Wei, Wang

    2009-01-01

    We present a hybrid cladding photonic crystal fiber for shaping high nonlinear and flattened dispersion in a wide range of wavelengths. The new structure adopts hybrid cladding with different pitches, air-holes diameters and air-holes arrayed fashions. The full-vector finite element method with perfectly matched layer is used to investigate the characteristics of the hybrid cladding photonic crystal fiber such as nonlinearity and dispersion properties. The influence of the cladding structure parameters on the nonlinear coefficient and geometric dispersion is analyzed. High nonlinear coefficient and the dispersion properties of fibers are tailored by adjusting the cladding structure parameters. A novel hybrid cladding photonic crystal fiber with high nonlinear coefficient and dispersion flattened which is suited for super continuum generation is designed. (author)

  8. Zirconium-barrier cladding attributes

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

    1987-01-01

    This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

  9. Laser cladding: repairing and manufacturing metal parts and tools

    Science.gov (United States)

    Sexton, Leo

    2003-03-01

    Laser cladding is presently used to repair high volume aerospace, automotive, marine, rail or general engineering components where excessive wear has occurred. It can also be used if a one-off high value component is either required or has been accidentally over-machined. The ultimate application of laser cladding is to build components up from nothing, using a laser cladding system and a 3D CAD drawing of the component. It is thus emerging that laser cladding can be classified as a special case of Rapid Prototyping (RP). Up to this point in time RP was seen, and is still seen, as in intermediately step between the design stage of a component and a finished working product. This can now be extended so that laser cladding makes RP a one-stop shop and the finished component is made from tool-steel or some alloy-base material. The marriage of laser cladding with RP is an interesting one and offers an alternative to traditional tool builders, re-manufacturers and injection mould design/repair industries. The aim of this paper is to discuss the emergence of this new technology, along with the transference of the process out of the laboratory and into the industrial workplace and show it is finding its rightful place in the manufacturing/repair sector. It will be shown that it can be used as a cost cutting, strategic material saver and consequently a green technology.

  10. 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

  11. 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

  12. 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)

  13. The study on radioactivity reduction of spent PWR cladding hull

    International Nuclear Information System (INIS)

    Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

    2003-01-01

    Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

  14. “Study of secondary hydriding at high temperature in zirconium based nuclear fuel cladding tubes by coupling information from neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and laser induced breakdown spectroscopy microprobe

    Energy Technology Data Exchange (ETDEWEB)

    Brachet, Jean-Christophe, E-mail: jean-christophe.brachet@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Hamon, Didier; Le Saux, Matthieu [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Vandenberghe, Valérie [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); DEN-Service d’Etudes Mécaniques et Thermiques (SEMT), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Toffolon-Masclet, Caroline; Rouesne, Elodie; Urvoy, Stéphane [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Béchade, Jean-Luc [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); DEN-Service de Recherches de Métallurgie Physique (SRMP), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Raepsaet, Caroline [LEEL, CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); NIMBE, CEA, CNRS, Université Paris-Saclay, 91191 Gif-sur-Yvette Cedex (France); and others

    2017-05-15

    This paper gives an overview of a multi-scale experimental study of the secondary hydriding phenomena that can occur in nuclear fuel cladding materials exposed to steam at high temperature (HT) after having burst (loss-of-coolant accident conditions). By coupling information from several facilities, including neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and micro laser induced breakdown spectroscopy, it was possible to map quantitatively, at different scales, the distribution of oxygen and hydrogen within M5™ clad segments having experienced ballooning and burst at HT followed by steam oxidation at 1100 and 1200 °C and final direct water quenching down to room temperature. The results were very reproducible and it was confirmed that internal oxidation and secondary hydriding at HT of a cladding after burst can lead to strong axial and azimuthal gradients of hydrogen and oxygen concentrations, reaching 3000–4000 wt ppm and 1.0–1.2 wt% respectively within the β phase layer for the investigated conditions. Consistent with thermodynamic and kinetics considerations, oxygen diffusion into the prior-β layer was enhanced in the regions highly enriched in hydrogen, where the α(O) phase layer is thinner and the prior-β layer thicker. Finally the induced post-quenching hardening of the prior-β layer was mainly related to the local oxygen enrichment. Hardening directly induced by hydrogen was much less significant. - Highlights: •More than 50% of the gaseous hydrogen produced by the inner clad oxidation absorbed and trapped into prior-β layer. •High hydrogen and oxygen local concentrations, up to 3000–4000 wt. ppm and 1.0–1.2 wt.% respectively, within the β phase. •Enhanced oxygen diffusion into hydrogen enriched prior-β layer, with locally thinner α(O) and thicker prior-β layers. •Post-quenching hardening of the prior-β structure mainly related to the (local) oxygen concentration.

  15. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

  16. YAG laser cladding to heat exchanger flange in actual plant

    International Nuclear Information System (INIS)

    Toshio, Kojima

    2001-01-01

    This paper is a sequel to ''Development of YAG Laser Cladding Technology to Heat Exchanger Flange'' presented in ICONE-8. A YAG Laser cladding technology is a permanent repairing and preventive maintenance method for heat exchanger's flange (channel side) seating surface which is degraded by the corrosion in long term operation. The material of this flange is carbon steel, and that of cladding wire is type 316 stainless steel so as to have high corrosion resistance. In former paper above, the soundness of cladding layers were presented to be verified. This channel side flange is bolted with tube sheet (shell side) through metal gasket. As the tube sheet side is already cladded a corrosion resistant material, it needs to apply the repairing and preventive maintenance method to only channel side. In 2000 this technology had been performed to the actual heat exchanger (Residual Heat Removal Heat Exchanger; RHR Hx) flange in domestic nuclear power plant. This paper described the outline, special equipment, and our total evaluation for this actual laser cladding work. And also several technical subjects which we should solve and/or improve for the next project was presented. (author)

  17. Hollow cylindrical plasma filament waveguide with discontinuous finite thickness cladding

    International Nuclear Information System (INIS)

    Alshershby, Mostafa; Hao Zuoqiang; Lin Jingquan

    2013-01-01

    We have explored here a hollow cylindrical laser plasma multifilament waveguide with discontinuous finite thickness claddi