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Sample records for reactor cladding applications

  1. 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)

  2. 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.

  3. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  4. Modification of OCA-I for application to a reactor pressure vessel with cladding on the inner surface

    International Nuclear Information System (INIS)

    Sauter, A.; Cheverton, R.D.; Iskander, S.K.

    1983-01-01

    The computer code OCA-I calculates the temperature distribution through the walls of a cylinder during a thermal transient and then performs a two-dimensional linear-elastic fracture-mechanics analysis to obtain stress-intensity factors for long surface flaws, considering both pressure and thermal loads. The code has been particularly useful in evaluating flaw behavior in reactor pressure vessels during overcooling accidents, but it has not previously treated the stainless steel cladding on the inner surface of the vessel as a discrete region. Although the cladding is quite thin compared with the base material, the large difference in thermal conductivity and coefficient of thermal expansion between the two materials results in a significant effect of the cladding on stress-intensity factors for surface cracks. Thus, the cladding was recently included as a discrete region in OCA-I

  5. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    2000-01-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  6. Development of austenitic stainless steel tubes for nuclear reactor cladding

    International Nuclear Information System (INIS)

    Padilha, A.F.; Ferreira, P.I.; Andrade, P.I.; Andrade, A.H.P. de; Meyerhof, S.; Mauricio, J.

    1984-01-01

    In the development of thin wall tubes for nuclear reactor fuel cladding applications, a great number of activities, related to the fabrication process as the qualification are involved. A test program was envisaged to verify the quality of seam welded stainless steel tubes (AISI 304), obtained as a result of an effort by the IPEN-CNEN/SP and the brazilian industry. The relevant aspects involved in the preparation of the tubes and some preliminary test results are presented. (Author) [pt

  7. 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

  8. 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)

  9. 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)

  10. Cladding and Duct Materials for Advanced Nuclear Recycle Reactors

    International Nuclear Information System (INIS)

    Allen, Todd R.; Busby, J. T.; Klueh, R. L.; Maloy, Stuart A.; Toloczko, Mychailo B.

    2008-01-01

    This is a review article that provides an overview of the reactor core structural materials and clad and duct needs for the GNEP advanced burner reactor design. A short history of previous research on structural materials for irradiation environments is provided. There is also a section describing some advanced materials that may be candidate materials for various reactor core structures

  11. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current US innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery

  12. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current U.S. innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery. (orig.)

  13. Coatings and claddings for the reduction of plasma contamination and surface erosion in fusion reactors

    International Nuclear Information System (INIS)

    Kaminsky, M.

    1980-01-01

    For the successful operation of plasma devices and future fusion reactors it is necessary to control plasma impurity release and surface erosion. Effective methods to obtain such controls include the application of protective coatings to, and the use of clad materials for, certain first wall components. Major features of the development programs for coatings and claddings for fusion applications will be described together with an outline of the testing program. A discussion of some pertinent test results will be included

  14. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    Kohn, E.; Wright, M.G.

    1976-11-01

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  15. Qualification of the LF-eddy current technique for the inspection of stainless steel cladding and applications on the reactor pressure vessel

    International Nuclear Information System (INIS)

    Weiss, R.; Becker, R.; Lucht, B.; Mohr, F.; Hartwig, K.

    2001-01-01

    As part of the re-inspection of the reactor pressure vessel of the nuclear power plant, the low-frequency-eddy current technique was implemented during the 1995 outage. Since then, this inspection technique and the testing equipment have seen steady further development. Therefore, optimization of the entire testing system, including qualification based on the 1995 results, was conducted. The eddy current testing system was designed as a ten-channel test system with sensors having separate transmitter and receiver coils. The first qualification of the testing technique and sensors was performed using a single-channel system; a second qualification was then carried out using the new testing electronics. The sensor design allows for a simultaneous detection of surface and subsurface flaws. This assumes that testing is performed simultaneously using four frequencies. Data analysis and evaluation are performed using a digital multi-frequency regression analysis technique The detection limits determined using this technique led to the definition of the following recording limits for testing in which the required signal-to-noise ratio of 6 dB was reliably observed. - Detection of surface connected longitudinal and transverse flaws: - notch, 3 mm deep and 10 mm long, for weave bead cladding; - notch, 2 mm deep and 20 mm long, for strip weld cladding. - Detection of embedded planar longitudinal and transverse flaws: - ligament of 7 mm for 8 mm clad thickness and 3 mm; - ligament for 4 mm clad thickness, notch starting at the carbon steel base material with a length of 20 mm. - Detection of embedded volumetric longitudinal and transverse flaws: - 3 mm diameter side-drilled hole (SDH) for 8 mm clad thickness; ligament, 4 mm. For 4 mm clad thickness: diameter, 2 mm SDH; ligament, 2 mm. All SDHs are 55 mm deep

  16. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  17. 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.

  18. 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.)

  19. NDT studies of laser cladding defects of pure copper on SS316L for in vessel materials for fusion reactor applications

    International Nuclear Information System (INIS)

    Shaikh, S.; Buddu, Ramesh Kumar; Raole, P.M.; Sarkar, B.

    2015-01-01

    The pure thick copper coatings of 1-3 mm are required for the in-vessel materials for the plasma facing components in fusion reactor systems to extract the very high heat flux in shorter durations (like VDEs) and to protect the in vessel components. Laser cladding technique is one of the potential technique for thick coatings on substrate materials. The present study reports the NDT characterization studies carried on samples of pure copper powder cladded on SS316L substrates of thickness 1 mm - 3 mm , fabricated by CO_2 laser system. Process parameters optimization like laser power, laser travel speed, spot size, powder feed rate and shield gas flow show the effect on quality of final cladding on steel substrates. X-ray radiography and Ultrasonic testing has been carried out thoroughly on the fabricated samples and defects are analyzed. Ultrasonic scan tests using different probes are employed as the interface defects are not thoroughly revealed by radiography. The calibration has been carried out by the test sample plate with known defect size created and various process parameters like amplitude, gain and metal velocity, relevant to specimen are chosen for probes calibration. The interface defects of porosity, lack of penetration, cracks or group porosities are observed in few set of samples developed. Radiography examination revealed the porosity at extreme edges and distributed porosity in the middle for thick cladding. Ultrasonic manual A-scanning with TR probe provides qualitative information about flaw and broadly gives its location of the defects. Samples of 1 mm thick cladding have shown relatively less porosity defects at the interface compared to 3 mm thick samples. (author)

  20. Oxidation during reflood of reactor core with melting cladding

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.; Allison, C.M.; Davis, K.L. [and others

    1995-09-01

    Models were recently developed and incorporated into the SCDAP/RELAP5 code for calculating the oxidation of fuel rods during cladding meltdown and reflood. Experiments have shown that a period of intense oxidation may occur when a hot partially oxidized reactor core is reflooded. This paper offers an explanation of the cladding meltdown and oxidation processes that cause this intense period of oxidation. Models for the cladding meltdown and oxidation processes are developed. The models are assessed by simulating a severe fuel damage experiment that involved reflood. The models for cladding meltdown and oxidation were found to improve calculation of the temperature and oxidation of fuel rods during the period in which hot fuel rods are reflooded.

  1. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  2. 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.

  3. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Bertram, W.

    1975-01-01

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  4. PCI resistant light water reactor fuel cladding

    International Nuclear Information System (INIS)

    Foster, J.P.; Sabol, G.P.

    1988-01-01

    A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities

  5. Solution treatment of fast reactor claddings

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro

    1974-01-01

    The fuel cladding tubes for Joyo (experimental FBR) are required to be a material corresponding to AISI Type 316 and cold-rolled after solution treatment. It is necessary to have no precipitation of carbide and to make the grain size smaller than ASTM No.6. It is very difficult to obtain the fine grains without the precipitation, however. In this connection, the behavior of carbide solution at high temperature and the annealing behavior of the material cold-worked and solution-treated were studied. The following matters are described: the solid solubility line of AISI Type 316; the behavior of carbide at solution treatment temperature; and the annealing behavior of the cold-worked material. (Mori, K.)

  6. Evolution of processing of GE fuel clad tubing for corrosion resistance in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Williams, C.D. [GE Nuclear Energy, Wilmington, NC (United States); Adamson, R.B. [GE Nuclear Energy, Wilmington, NC (United States); Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Plaza-Meyer, E. [GE Nuclear Energy, Wilmington, NC (United States); Proebstle, R.A. [GE Nuclear Energy, Wilmington, NC (United States); White, D.W. [GE Nuclear Energy, Wilmington, NC (United States)

    1996-05-01

    The current modification of the primary GE in-process solution-quench heat treatment, an (alpha+beta) solution-quench carried out at a tube diameter requiring only two subsequent reduction and anneal cycles, is applicable to Zr barrier fuel clad tubing, to non-barrier fuel clad tubing, and to the TRICLAD tubing product. A combination of good in-reactor corrosion performance and degradation resistance is anticipated for these products, based on knowledge of metallurgical characteristics and supported by the demonstrated performance capability of the Zircaloy-2 materials used. (orig.)

  7. Impacts of reactor. Induced cladding defects on spent fuel storage

    International Nuclear Information System (INIS)

    Johnson, A.B.

    1978-01-01

    Defects arise in the fuel cladding on a small fraction of fuel rods during irradiation in water-cooled power reactors. Defects from mechanical damage in fuel handling and shipping have been almost negligible. No commercial water reactor fuel has yet been observed to develop defects while stored in spent fuel pools. In some pools, defective fuel is placed in closed canisters as it is removed from the reactor. However, hundreds of defective fuel bundles are stored in numerous pools on the same basis as intact fuel. Radioactive species carried into the pool from the reactor coolant must be dealt with by the pool purification system. However, additional radiation releases from the defective fuel during storage appear tu be minimal, with the possible exception of fuel discharged while the reactor is operating (CANDU fuel). Over approximately two decades, defective commercial fuel has been handled, stored, shipped and reprocessed. (author)

  8. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  9. 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

  10. Occurence and prediction of sigma phase in fuel cladding alloys for breeder reactors

    International Nuclear Information System (INIS)

    Anantatmula, R.P.

    1982-01-01

    In sodium-cooled fast reactor systems, fuel cladding materials will be exposed for several thousand hours to liquid sodium. Satisfactory performance of the materials depends in part on the sodium compatibility and phase stability of the materials. This paper mainly deals with the phase stability aspect, with particular emphasis on sigma phase formation of the cladding materials upon extended exposures to liquid sodium. A new method of predicting sigma phase formation is proposed for austenitic stainless steels and predictions are compared with the experimental results on fuel cladding materials. Excellent agreement is obtained between theory and experiment. The new method is different from the empirical methods suggested for superalloys and does not suffer from the same drawbacks. The present method uses the Fe-Cr-Ni ternary phase diagram for predicting the sigma-forming tendencies and exhibits a wide range of applicability to austenitic stainless steels and heat-resistant Fe-Cr-Ni alloys

  11. Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor

    International Nuclear Information System (INIS)

    Chen Daolong

    1986-01-01

    A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements

  12. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  13. Properties of light water reactor spent fuel cladding. Interim report

    International Nuclear Information System (INIS)

    Farwick, D.G.; Moen, R.A.

    1979-08-01

    The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

  14. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2003-01-01

    This report describes research performed in ten laboratories within the framework of the IAEA Co-ordinated Research Project on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water. The project consisted of exposure of standard racks of corrosion coupons in the spent fuel pools of the participating research reactor laboratories and the evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage water. A group of experts in the field contributed a state of the art review and provided technical supervision of the project. Localized corrosion mechanisms are notoriously difficult to understand, and it was clear from the outset that obtaining consistency in the results and their interpretation from laboratory to laboratory would depend on the development of an excellent set of experimental protocols. These experimental protocols are described in the report together with guidelines for the maintenance of optimum water chemistry to minimize the corrosion of aluminium clad research reactor fuel in wet storage. A large database on corrosion of aluminium clad materials has been generated from the CRP and the SRS corrosion surveillance programme. An evaluation of these data indicates that the most important factors contributing to the corrosion of the aluminium are: (1) High water conductivity (100-200 μS/cm); (2) Aggressive impurity ion concentrations (Cl - ); (3) Deposition of cathodic particles on aluminium (Fe, etc.); (4) Sludge (containing Fe, Cl - and other ions in concentrations greater than ten times the concentrations in the water); (5) Galvanic couples between dissimilar metals (stainless steel-aluminium, aluminium-uranium, etc); (6) Scratches and imperfections (in protective oxide coating on cladding); (7) Poor water circulation. These factors operating both independently and synergistically may cause corrosion of the aluminium. The single most important key to preventing corrosion is maintaining good

  15. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  16. Evaluation of the applicability of cladding deformation model in RELAP5/MOD3.2 code for VVER-1000 fuel

    International Nuclear Information System (INIS)

    Vorob'ev, Yu.; Zhabin, O.

    2015-01-01

    Applicability of cladding deformation model in RELAP5/MOD3.2 code is analyzed for VVER-1000 fuel cladding from Zr+1%Nb alloy. Experimental data and calculation model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the cladding temperature range from 600 to 1200 deg C and pressure range from 1 to 12 MPa. Evaluation results demonstrate limited applicability of built-in RELAP5/MOD3.2 cladding deformation model to the estimation of Zr+1%Nb cladding rupture conditions. The limitations found shall be considered in application of RELAP5/MOD3.2 cladding deformation model in the design-basis accident analysis of VVER reactors

  17. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  18. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  19. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2008-08-31

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  20. Corrosion effect of fast reactor fuel claddings on their mechanical properties

    International Nuclear Information System (INIS)

    Davydov, E.F.; Krykov, F.N.; Shamardin, V.K.

    1985-01-01

    Fast reactor fuel cladding corrosion effect on its mechanical properties was investigated. UO 2 fuel elements were irradiated in the BOP-60 reactor at the linear heat rate of 42 kw/m. Fuel cladding is made of stainless steel OKh16N15M3BR. Calculated maximum cladding temperature is 920 K. Neutron fluence in the central part of fuel elements is 6.3x10 26 m+H- 2 . To investigate the strength changes temperature dependence of corrossion depth, cladding strength reduction factors was determined. Samples plasticity reduction with corrosion layer increase is considered to be a characteristic feature

  1. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    1987-09-01

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

  2. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

    1977-04-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  3. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    International Nuclear Information System (INIS)

    Boltax, A.; Biancheria, A.

    1977-01-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  4. Fabrication of the fuel elements cladding for utilization in the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Schaeffer, L.; Sefidvash, F.

    1986-01-01

    A method for the fabrication of cladding of the spherical fuel elements for the utilization in the fluidized bed nuclear reactor is presented. Some prelimminary experiments were performed to adopt a method which adapt itself to mass production with the desired high quality. Still methods for cladding fabrication are under study. (Author) [pt

  5. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-01

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean

  6. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  7. Cladding modes of optical fibers: properties and applications

    International Nuclear Information System (INIS)

    Ivanov, Oleg V; Nikitov, Sergei A; Gulyaev, Yurii V

    2006-01-01

    One of the new methods of fiber optics uses cladding modes for controlling propagation of radiation in optical fibers. This paper reviews the results of studies on the propagation, excitation, and interaction of cladding modes in optical fibers. The resonance between core and cladding modes excited by means of fiber Bragg gratings, including tilted ones, is analyzed. Propagation of cladding modes in microstructured fibers is considered. The most frequently used method of exciting cladding modes is described, based on the application of long-period fiber gratings. Examples are presented of long-period gratings used as sensors and gain equalizers for fiber amplifiers, as well as devices for coupling light into and out of optical fibers. (instruments and methods of investigation)

  8. Observations of in-reactor endurance and rupture life for fueled and unfueled FTR cladding

    International Nuclear Information System (INIS)

    Lovell, A.J.; Christensen, B.Y.; Chin, B.A.

    1979-01-01

    Reactor component endurance limits are important to nuclear experimenters and operators. This paper investigates endurance limits of 316 CW fuel pin cladding. The objective of this paper is to compare and analyze two different sets of FTR fuel pin cladding data. The first data set is from unfueled pressurized cladding irradiated in the Experimental Breeder Reactor No. II (EBR-II). This data set was generated in an assembly in which the temperature was monitored and controlled. The second data set contains observations of breached and unbreached EBR-II test fuel pins covering a large range of temperature, power and burnup conditions

  9. Development of ODS (oxide dispersion strengthened) ferritic-martensitic steels for fast reactor fuel cladding

    International Nuclear Information System (INIS)

    Ukai, Shigeharu

    2000-01-01

    In order to attain higher burnup and higher coolant outlet temperature in fast reactor, oxide dispersion strengthened (ODS) ferritic-martensitic steels were developed as a long life fuel cladding. The improvement in formability and ductility, which are indispensable in the cold-rolling method for manufacturing cladding tube, were achieved by controlling the microstructure using techniques such as recrystallization heat-treatment and α to γ phase transformation. The ODS ferritic-martensitic cladding tubes manufactured using these techniques have the highest internal creep rupture strength in the world as ferritic stainless steels. Strength level approaches adequate value at 700degC, which meets the requirement for commercial fast reactors. (author)

  10. 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)

  11. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  12. Pellet-clad interaction in water reactor fuels

    International Nuclear Information System (INIS)

    2004-01-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  13. 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.

  14. Alloy development for cladding and duct applications

    International Nuclear Information System (INIS)

    Straalsund, J.L.; Johnson, G.D.

    1981-01-01

    Three general classes of materials under development for cladding and ducts are listed. Solid solution strengthened, or austenitic, alloys are Type 316 stainless steel and D9. Precipitation hardened (also austenitic) alloys consist of D21, D66 and D68. These alloys are similar to such commercial alloys as M-813, Inconel 706, Inconel 718 and Nimonic PE-16. The third general class of alloys is composed of ferritic alloys, with current emphasis being placed on HT-9, a tempered martensitic alloy, and D67, a delta-ferritic steel. The program is comprised of three parallel paths. The current reference, or first generation alloy, is 20% cold worked Type 316 stainless steel. Second generation alloys for near-term applications include D9 and HT-9. Third generation materials consist of the precipitation strengthened steels and ferritic alloys, and are being considered for implementation at a later time than the first and second generation alloys. The development of second and third generation materials was initiated in 1974 with the selection of 35 alloys. This program has proceeded to today where there are six advanced alloys being evaluated. These alloys are the developmental alloys D9, D21, D57, D66 and D68, together with the commerical alloy, HT-9. The status of development of these alloys is summarized

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  16. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

    International Nuclear Information System (INIS)

    Alencar, Donizete Anderson de

    2004-01-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  17. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  18. Welding of stainless steel clad fuel rods for nuclear reactors

    International Nuclear Information System (INIS)

    Neves, Mauricio David Martins das

    1986-01-01

    This work describes the obtainment of austenitic stainless steel clad fuel rods for nuclear reactors. Two aspects have been emphasized: (a) obtainment and qualification of AISI 304 and 304 L stainless steel tubes; b) the circumferential welding of pipe ends to end plugs of the same alloy followed by qualification of the welds. Tubes with special and characteristic dimensions were obtained by set mandrel drawing. Both, seamed and seamless tubes of 304 and 304 L were obtained.The dimensional accuracy, surface roughness, mechanical properties and microstructural characteristics of the tubes were found to be adequate. The differences in the properties of the tubes with and without seams were found to be insignificant. The TIG process of welding was used. The influence of various welding parameters were studied: shielding gas (argon and helium), welding current, tube rotation speed, arc length, electrode position and gas flow. An inert gas welding chamber was developed and constructed with the aim of reducing surface oxidation and the heat affected zone. The welds were evaluated with the aid of destructive tests (burst-test, microhardness profile determination and metallographic analysis) and non destructive tests (visual inspection, dimensional examination, radiography and helium leak detection). As a function of the results obtained, two different welding cycles have been suggested; one for argon and another for helium. The changes in the microstructure caused by welding have been studied in greater detail. The utilization of work hardened tubes, permitted the identification by optical microscopy and microhardness measurements, of the different zones: weld zone; heat affected zone (region of grain growth, region of total and partial recrystallization) and finally, the zone not affected by heat. Some correlations between the welding parameters and metallurgical phenomena such as: solidification, recovery, recrystallization, grain growth and precipitation that occurred

  19. Some aspects of the utilization of zicaloy and austenitic steel as cladding material for PWR reactor fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Perrotta, J.A.

    1985-01-01

    The behaviour under irradiation of fuel rods for light water reactors was simulated by using fuel performance codes. Two types of cladding were analyzed: zircaloy and austenitic stainless steel. The fuel performance codes, originally made for zircaloy cladding, were adapted for austenitic stainless steel. The simulation results for the two types of cladding are presented, compared and discussed. (F.E.) [pt

  20. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  1. Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

    Science.gov (United States)

    Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

    2018-05-01

    The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (steel cladding is retained despite He2+ implantation.

  2. In-reactor measurement of clad strain: effect of power history

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Morel, P.A.

    1980-01-01

    A series of experimental irradiations has been undertaken at CRNL to measure directly the in-reactor deformation of fuel elements while they are operating at power. Power histories have been chosen to allow investigation of power, time at power and burnup on pellet-clad interaction for element linear powers to 60kW/m. Results are presented which indicate that irradiation of a fresh fuel element at high power is effective in minimizing clad hoop stresses during subsequent ramps or cycles to that power. The effectiveness of this preconditioning appears to be due primarily to fuel densification rather than stress relaxation in the clad. (auth)

  3. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  4. Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Kapoor, K.; Prahlad, B.

    2012-01-01

    Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

  5. Pellet-clad interaction observations in boiling water reactor fuel elements

    International Nuclear Information System (INIS)

    Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

    1981-01-01

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  6. Process for surface treatment of zirconium-containing cladding materials for fuel element or other components for nuclear reactors

    International Nuclear Information System (INIS)

    Videm, K.G.; Lunde, L.R.; Kooyman, H.H.

    1975-01-01

    A process for the surface treatment of zirconium-base cladding materials for fuel elements or other components for nuclear reactors is described. The treatment includes pickling the cladding material in a fluoride-containing bath, and then applying a protective coating through oxidation to the pickled cladding material. The fluoride-containing contaminants which remain on the surface of the cladding material during pickling are removed or rendered harmless by anodic oxidation

  7. Application of YAG laser cladding to the flange seating surface

    International Nuclear Information System (INIS)

    Nakanishi, Koki; Ninomiya, Kazuyuki; Nezaki, Koji

    1999-01-01

    Stainless cladding on carbon steel is usually conducted by shielded metal arc welding (SMAW) or gas tungsten arc welding (GTAW). YAG ( Yttrium-Aluminum-Garnet) laser welding is superior to these methods of welding in the following respects : (1) The heat affected zone (HAZ) is narrower and there is less distortion. (2) YAG laser cladding has the required chemical compositions, even with possibly fewer welding layers under controlled dilution. (3) Greater welding speed. YAG laser cladding application to vessel flange seating surfaces was examined in this study and the results are discussed. The following objectives were carried out : (1) Determination of welding conditions for satisfactory cladding layers and (2) whether cladding would be adequately possible at a cornered section of a stair-like plate, assuming actual flange shape. (3) Measurement of welding distortion and heat affected zone in carbon steel. The welding conditions for producing no-crack deposit with low dilution in carbon steel were clarified and welding by which cladding at cornered section would be possible was achieved. welding distortion by YAG laser was found less than with GTAW and HAZ made by first layer welding could be tempered appropriately by second layer welding. (author)

  8. Fuel cladding tube and fuel rod for BWR type reactor

    International Nuclear Information System (INIS)

    Urata, Megumu; Mitani, Shinji.

    1995-01-01

    A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

  9. Cladding tube of fuel rod for a BWR type reactor

    International Nuclear Information System (INIS)

    Nakayama, Hitoshi; Fujie, Kunio; Kuwahara, Heikichi; Hirai, Tadamasa; Kakizaki, Kimio.

    1976-01-01

    Object: To form a cladding tube wall with tunnels in communication with the exterior through a number of small-diameter openings to rapidly disperse a large quantity of heat thereby providing high density of the fuel rod. Structure: Tunnels adjacent to each other are provided under the skin in contact with cooling liquid of a cladding tube, and a number of openings through which said tunnels and the periphery of the cladding tube are placed in communication are formed, said openings each having its section smaller than that of said tunnel. With this arrangement, the cooling water entered the tunnel through some of small diameter openings absorbs heat of the fuel rod to be vaporized, which is flown out into the cooling water through the other small diameter openings and formed into vapor bubbles which move up for release of heat. (Taniai, N.)

  10. Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.

    1978-06-01

    Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

  11. Effect of reactor chemistry and operating variables on fuel cladding corrosion in PWRs

    International Nuclear Information System (INIS)

    Park, Moon Ghu; Lee, Sang Hee

    1997-01-01

    As the nuclear industry extends the fuel cycle length, waterside corrosion of zircaloy cladding has become a limiting factor in PWR fuel design. Many plant chemistry factors such as, higher lithium/boron concentration in the primary coolant can influence the corrosion behavior of zircaloy cladding. The chemistry effect can be amplified in higher duty fuel, particularlywhen surface boiling occurs. Local boiling can result in increased crud deposition on fuel cladding which may induce axial power offset anomalies (AOA), recently reported in several PWR units. In this study, the effect of reactor chemistry and operating variables on Zircaloy cladding corrosion is investigated and simulation studies are performed to evaluate the optimal primary chemistry condition for extended cycle operation. (author). 8 refs., 3 tabs., 16 figs

  12. Circumferential nonuniformity of cladding radiation swelling of fast reactor peripheral fuel elements

    International Nuclear Information System (INIS)

    Reutov, V.F.; Farkhutdinov, K.G.

    1977-01-01

    The results are presented of the investigation into the perimeter radiation swelling of Kh18N10T stainless steel cladding in different cross sections of a peripheral fuel element of the BR-5 reactor. The fluence on the cladding is 1.8-2.9 x 10 22 fast neutr/cm 2 , the operating temperatures in different parts of the fuel element being 430 deg to 585 deg C. There has been observed circumferential non-uniformity of the distribution, concentration, and of the total volume of radiation cavities, which is due to temperature non-uniformity along the cladding perimeter. It is shown that such non-uniformity of radiation swelling of the cladding material may result in bending of the peripheral fuel element with regard to the fuel assembly sheath walls

  13. Design optimization of multi-layer Silicon Carbide cladding for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@unm.edu [Department of Nuclear Engineering, University of New Mexico, MSC01 1120 1 University of New Mexico, Albuquerque, NM 87131 (United States); NO, Hee Cheon, E-mail: hcno@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2017-01-15

    Highlights: • SiC cladding designs are optimized with a multi-layer structural analysis code. • Layer radial thickness fraction that minimizes cladding fracture probability exists. • The demonstrated procedure is applicable for multi-layer SiC cladding design. • Duplex SiC with the inner composite fraction ∼0.4 is optimal in a reference case. • Increasing composite thermal conductivity markedly decreases SiC cladding stress. - Abstract: A parametric study that demonstrates a methodology for determining the optimum bilayer composition in a duplex SiC cladding is discussed. The structural performance of multi-layer SiC cladding design is significantly affected by radial thickness fraction of each layer. This study shows that there exists an optimal composite/monolith radial thickness fraction that minimizes failure probability for a duplex SiC cladding in steady-state operation. An exemplary reference case study shows that the duplex cladding with the inner composite fraction ∼0.4 and the outer CVD-SiC fraction ∼0.6 is found to be the optimal SiC cladding design for the current PWRs with the reference material choice for CVD-SiC and fiber reinforced composite. A marginal increase in the composite fraction from the presented optimal designs may lead to increase structural integrity by introducing some unquantified merits such as increasing damage tolerance. The major factors that affect the optimum cladding designs are temperature gradients and internal gas pressure. Clad wall thickness, thermal conductivity, and Weibull modulus are among the key design parameters/material properties.

  14. Development and fabrication of seamless Aluminium finned clad tubes for metallic uranium fuel rods for research reactor

    International Nuclear Information System (INIS)

    Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.

    2012-01-01

    Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results

  15. 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.

  16. Corrosion of research reactor aluminium clad spent fuel in water. Additional information

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  17. Theoretical investigations of the meltoff and resolidification process of fuel claddings during accidents in liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Angerer, G.

    1978-08-01

    During loss-of-coolant-flow accidents in liquid metal cooled fast breeder reactors with failure to scram the fuel claddings will melt after boiling and evaporation of the coolant. The CMOT model presented here describes the subsequent process of relocation and resolidification of the molten claddings. The basic thermohydrodynamics equations of the two-phase flow of cladding material and sodium vapor are solved numerically by differential approximations in a Eulerian reference net. The results calculated by the model improved the insight into the dynamics of the cladding relocation process. Here are the main results: - Shortly after the onset of cladding relocation large waves of molten cladding material are generated. The motion of these waves contributes considerably to the material transport. - The dynamics of cladding relocation exhibits strong local incoherences. - The formation of cladding blockages observed at the ends of the fuel region is confirmed by the calculations. - In case of incoherent cladding meltoff less cladding material is transported upwards. - Cladding relocation strongly depends on the axial pressure drop and the underlying friction factor correlations. Recalculation of the R5 loss-of-coolant-flow experiment performed in the U.S. TREAT test reactor is in good agreement with the experimental data. (orig./HP) 891 HP [de

  18. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  19. Protection of spent aluminum-clad research reactor fuels during extended wet storage

    International Nuclear Information System (INIS)

    Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

    2013-01-01

    Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

  20. 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

  1. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

  2. Corrosion and protection of spent Al-clad research reactor fuel during extended wet storage

    International Nuclear Information System (INIS)

    Ramanathan, Lalgudi V.

    2009-01-01

    A variety of spent research reactor fuel elements with different fuel meats, geometries and 235 U enrichments are presently stored under water in basins throughout the world. More than 90% of these fuels are clad in aluminum (Al) or its alloy and are susceptible to corrosion. This paper presents an overview of the influence of Al alloy composition, galvanic effects (Al alloy/stainless steel), crevice effects, water parameters and synergism between these parameters as well as settled solids on the corrosion of typical Al alloys used as fuel element cladding. Pitting is the main form of corrosion and is affected by water conductivity, chloride ion content, formation of galvanic couples with rack supports and settled solid particles. The extent to which these parameters influence Al corrosion varies. This paper also presents potential conversion coatings to protect the spent fuel cladding. (author)

  3. In-reactor cladding breach of EBR-II driver-fuel elements

    International Nuclear Information System (INIS)

    Seidel, B.R.; Einziger, R.E.

    1977-01-01

    Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

  4. 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

  5. Microstructural investigations of fast reactor irradiated austenitic and ferritic-martensitic stainless steel fuel cladding

    International Nuclear Information System (INIS)

    Agueev, V.S.; Medvedeva, E.A.; Mitrofanova, N.M.; Romanueev, V.V.; Tselishev, A.V.

    1992-01-01

    Electron microscopy has been used to characterize the microstructural changes induced in advanced fast reactor fuel claddings fabricated from Cr16Ni15Mo3NbB and Cr16Ni15Mo2Mn2TiVB austenitic stainless steels in the cold worked condition and Cr13Mo2NbVB ferritic -martensitic steel following irradiation in the BOR-60, BN-350 and BN-600 fast reactors. The data are compared with the results obtained from a typical austenitic commercial cladding material, Cr16Ni15Mo3Nb, in the cold worked condition. The results reveal a beneficial effect of boron and other alloying elements in reducing void swelling in 16Cr-15Ni type austenitic steels. The high resistance of ferritic-martensitic steels to void swelling has been confirmed in the Cr13Mo2NbVB steel. (author)

  6. Flaw density examinations of a clad boiling water reactor pressure vessel segment

    International Nuclear Information System (INIS)

    Cook, K.V.; McClung, R.W.

    1986-01-01

    Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel [nominally 0.7 by 3 m (2 by 10 ft)]. This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected

  7. Chemical aspects of pellet-cladding interaction in light water reactor fuel elements

    International Nuclear Information System (INIS)

    Olander, D.R.

    1982-01-01

    In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI

  8. Development of a laser multi-layer cladding technology for damage mitigation of fuel spacers in Hanaro reactor

    International Nuclear Information System (INIS)

    Kim, J. S.; Lee, D. H.; Hwang, S. S.; Suh, J. H.

    2002-01-01

    A laser multi-layer cladding technology was developed to mitigate the fretting wear damages occurred at fuel spacers in Hanaro reactor. The detailed experimental results are as follows. 1) Analyses of fretting wear damages and fabrication process of fuel spacers 2) Development and analysis of spherical Al 6061 T-6 alloy powders for the laser cladding 3) Analysis of parameter effects on laser cladding process for clad bids, and optimization of laser cladding process 4) Analysis on the changes of cladding layers due to overlapping factor change 5) Microstructural observation and phase analysis 6) Characterization of materials properties (hardness and wear tests) 7) Manufacture of prototype fuel spacers 8) Development of a vision system and revision of its related softwares

  9. Non-destructive control of cladding thickness of fuel elements for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karlov, Y.; Zhukov, Y.; Chashchin, S

    1997-07-01

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  10. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira; Giovedi, Claudia

    2015-01-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  11. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  12. The role of a fuel element and its cladding in water cooled reactor dynamics

    International Nuclear Information System (INIS)

    Randles, J.

    1963-10-01

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO 2 fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

  13. Current status of materials development of nuclear fuel cladding tubes for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

    2017-05-15

    Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

  14. Contribution to numerical and mechanical modelling of pellet-cladding interaction in nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Retel, V.

    2002-12-01

    Pressurised water reactor fuel rods (PWR) are the place of nuclear fission, resulting in unstable and radioactive elements. Today, the mechanical loading on the cladding is harder and harder and is partly due to the fuel pellet movement. Then, the mechanical behaviour of the cladding needs to be simulated with models allowing to assess realistic stress and strain fields for all the running conditions. Besides, the mechanical treatment of the fuel pellet needs to be improved. The study is part of a global way of improving the treatment of pellet-cladding interaction (PCI) in the 1D finite elements EDF code named CYRANO3. Non-axisymmetrical multidirectional effects have to be accounted for in a context of unidirectional axisymmetrical finite elements. The aim of this work is double. Firstly a model simulating the effect of stress concentration on the cladding, due to the opening of the radial cracks of fuel, had been added in the code. Then, the fragmented state of fuel material has been taken into account in the thermomechanical calculation, through a model which led the strain and stress relaxation in the pellet due to the fragmentation, be simulated. This model has been implemented in the code for two types of fuel behaviour: elastic and viscoplastic. (author)

  15. The role of a fuel element and its cladding in water cooled reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1963-10-15

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO{sub 2} fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

  16. Current status of materials development of nuclear fuel cladding tubes for light water reactors

    International Nuclear Information System (INIS)

    Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie; Abe, Hiroaki

    2017-01-01

    Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

  17. In-reactor performance of methods to control fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G.

    1979-01-01

    Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (Δ anti GO 2 ) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (Δ anti GO 2 ) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins

  18. Temperature distribution determination of JPSR power reactor fuel element and cladding

    International Nuclear Information System (INIS)

    Sudarmono

    1996-01-01

    In order to utilize of fuel rod efficiency, a concept of JAERI passive Safety Reactor (JPSR) has been developed in Japan Atomic Energy Research Institute. In the JPSR design, UO 2 . are adopted as a fuel rod. The temperature distribution in the fuel rod and cladding in the hottest channel is a potential limiting design constraint of the JPSR. In the present determination, temperature distribution of the fuel rod and cladding for JPSR were PET:formed using COBRA-IV-I to evaluate the safety margin of the present JPSR design. In this method, the whole core was represented by the 1/4 sector and divided into 50 subchannels and 40 axial nodes. The temperature become maximum at the elevation of 1.922 and 2.196 m in the typical cell under operating condition. The maximum temperature in the center of the fuel rod surface of the fuel rod and cladding were 1620,4 o C, 722,8 o C, and 348,6 o C. The maximum results of temperature in the center of the fuel rod and cladding; were 2015,28 o C and 550 o C which were observed at 3.1 second in the typical cell

  19. Characterisation of the activities associated with fast reactor cladding hulls

    International Nuclear Information System (INIS)

    Biddle, P.; Rees, J.H.

    1987-01-01

    Four batches of hulls from the reprocessing of fuel from the Prototype Fast Reactor at Dounreay have been examined visually and photographically. The appearance of the hulls (shiny or black) depended on the original location in the fuel pin. The total fissile activities measured corresponded to less than 0.1% or the original fuel. The α-emitter, fission product and activation product activities have been determined from dissolution studies. Profile diagrams have been constructed for the distribution of the principal activities along lengths of seventeen hulls. The evidence obtained suggested that much of the activity on the hulls arose through deposition from the dissolver liquors. The deposits included insoluble particulate material which was thought to be responsible for the wide variations in activity across short lengths of some of the hulls. (orig.)

  20. Corrosion of aluminium-clad spent fuel at RA research reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Maksin, T.; Dobrijevic, R.; Idjakovic, Z.

    2003-01-01

    Almost 95% of all spent fuel elements of the RA research reactor in the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, are stored in 30 aluminium barrels and about 300 stainless steel channel-holders in the temporary spent fuel storage water pool. The first activities of sludge and water samples, taken from the pool, were measured in 1996-1997 and were followed by analysis of chemical composition of samples. Visual inspections of fuel elements in some stainless steel tubes and of the fuel channels stored in the reactor core have shown that some deposits cover aluminium cladding. Stains and surface discoloration are noted on many of the spent fuel elements that were examined visually during the core unloading and inspections carried out in 1979 - 1984. Some of water samples, taken from pool, about a 150 stainless steel tubes and 16 barrels have shown very high 137-Cs activity compared to low activity measured in pool water. It was concluded that aluminium cladding of the fuel elements was penetrated due to corrosion process. Study on influence of water corrosion processes in the RA reactor storage pool was started within the framework of the IAEA CRP 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water' in 2002. The first test rack with various aluminium and stainless steel coupons, supplied by the IAEA, was immersed in the pool already in 1996. New racks were immersed in 2002 and 2003. The rack immersed in 1996 was taken out from the pool in 2002 and the rack immersed in 2002 was taken out in 2003. Results of the examination of these racks, carried out according to the strategy and the protocol, proposed by the IAEA, are described in this paper. (author)

  1. Analysis of clad motion during a loss of flow (LOF) accident in a fast sodium cooled reactor

    International Nuclear Information System (INIS)

    Henkel, P.

    1985-10-01

    A new model describing clad motion during a Loss of Flow (LOF) accident in a Liquid Metal Cooled Fast (Breeder) Reactor (LMFBR) is presented. Its special features are Clad motion is treated within a fuel pin bundle. The bundle geometry is represented by an equivalent annular geometry which serves as the descriptional basis for the clad motion analysis; Several flow regimes are considered. These include a wave or film flow along the fuel pin surfaces as well as a drop flow within the coolant channels. A new entrainment criterion is successfully applied to describe the entrainment of molten cladding and the coolant flow is modelled as a two-dimensional, monstationary flow. Therefore, radial cross flows in a pin bundle can be calculated. Especially, thermal incoherency effects can be treated consistently. The analysis of clad motion in the two experiments STAR1 and STAR2 using the subsequently presented SANDCMOT model gives good agreement with the experimental data. (orig.) [de

  2. Corrosion of research reactor Al-clad spent fuel in water

    International Nuclear Information System (INIS)

    Bendereskaya, O.S.; De, P.K.; Haddad, R.; Howell, J.P.; Johnson, A.B. Jr.; Laoharojanaphand, S.; Luo, S.; Ramanathan, L.V.; Ritchie, I.; Hussain, N.; Vidowsky, I.; Yakovlev, V.

    2002-01-01

    A significant amount of aluminium-clad spent nuclear fuel from research and test reactors worldwide is currently being stored in water-filled basins while awaiting final disposition. As a result of corrosion issues, which developed from the long-term wet storage of aluminium-clad fuel, the International Atomic Energy Agency (IAEA) implemented a Co-ordinated Research Project (CRP) in 1996 on the 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water'. The investigations undertaken during the CRP involved ten institutes in nine different countries. The IAEA furnished corrosion surveillance racks with aluminium alloys generally used in the manufacture of the nuclear fuel cladding. The individual countries supplemented these racks with additional racks and coupons specific to materials in their storage basins. The racks were immersed in late 1996 in the storage basins with a wide range of water parameters, and the corrosion was monitored at periodic intervals. Results of these early observations were reported after 18 months at the second research co-ordination meeting (RCM) in Sao Paulo, Brazil. Pitting and crevice corrosion were the main forms of corrosion observed. Corrosion caused by deposition of iron and other particles on the coupon surfaces was also observed. Galvanic corrosion of stainless steel/aluminium coupled coupons and pitting corrosion caused by particle deposition was observed. Additional corrosion racks were provided to the CRP participants at the second RCM and were immersed in the individual basins by mid-1998. As in the first set of tests, water quality proved to be the key factor in controlling corrosion. The results from the second set of tests were presented at the third and final RCM held in Bangkok, Thailand in October 2000. An IAEA document giving details about this CRP and other guidelines for spent fuel storage is in pres. This paper presents some details about the CRP and the basis for its extension. (author)

  3. 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.

  4. A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tahk, Y. W.; Oh, J. Y.; Lee, B. H.; Seo, C. G.; Chae, H. T.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    U{sub 3}Si{sub 2}-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

  5. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Batte, G.L.; Pahl, R.G.; Hofman, G.L.

    1993-01-01

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II

  6. Use of ion beams to simulate reaction of reactor fuels with their cladding

    International Nuclear Information System (INIS)

    Birtcher, R.C.; Baldo, P.

    2006-01-01

    Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium-molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium-molybdenum coated with a 100 nm thin film of aluminum was irradiated with 3 MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27 Al(p,γ) 28 Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 deg. C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm 2 /dpa. At a dose of 90 dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery

  7. Fuel-to-cladding heat transfer coefficient into reactor fuel element

    International Nuclear Information System (INIS)

    Lassmann, K.

    1979-01-01

    Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

  8. Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

    Science.gov (United States)

    Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

    2018-02-01

    The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

  9. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

    Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

  10. A statistical analysis of pellet-clad interaction failures in water reactor fuel

    International Nuclear Information System (INIS)

    McDonald, S.G.; Fardo, R.D.; Sipush, P.J.; Kaiser, R.S.

    1981-01-01

    The primary objective of the statistical analysis was to develop a mathematical function that would predict PCI fuel rod failures as a function of the imposed operating conditions. Linear discriminant analysis of data from both test and commercial reactors was performed. The initial data base used encompassed 713 data points (117 failures and 596 non-failures) representing a wide variety of water cooled reactor fuel (PWR, BWR, CANDU, and SGHWR). When applied on a best-estimate basis, the resulting function simultaneously predicts approximately 80 percent of both the failure and non-failure data correctly. One of the most significant predictions of the analysis is that relatively large changes in power can be tolerated when the pre-ramp irradiation power is low, but that only small changes in power can be tolerated when the pre-ramp irradiation power is high. However, it is also predicted that fuel rods irradiated at low power will fail at lower final powers than those irradiated at high powers. Other results of the analysis are that fuel rods with high clad operating temperatures can withstand larger power increases that fuel rods with low clad operating temperatures, and that burnup has only a minimal effect on PCI performance after levels of approximately 10000 MWD/MTU have been exceeded. These trends in PCI performance and the operating parameters selected are believed to be consistent with mechanistic considerations. Published PCI data indicate that BWR fuel usually operates at higher local powers and changes in power, lower clad temperatures, and higher local ramp rates than PWR fuel

  11. Corrosion surveillance programme for Latin American research reactor Al-clad spent fuel in water

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Haddad, R.; Ritchie, I.

    2002-01-01

    The objectives of the IAEA sponsored Regional Technical Co-operation Project for Latin America (Argentina, Brazil, Chile, Mexico, and Peru) are to provide the basic conditions to define a regional strategy for managing spent fuel and to provide solutions, taking into consideration the economic and technological realities of the countries involved. In particular, to determine the basic conditions for managing research reactor spent fuel during operation and interim storage as well as final disposal, and to establish forms of regional cooperation in the four main areas: spent fuel characterization, safety, regulation and public communication. This paper reports the corrosion surveillance activities of the Regional Project and these are based on the IAEA sponsored co-ordinated research project (CRP) on 'Corrosion of research reactor Al-clad spent fuel in water'. The overall test consists of exposing corrosion coupon racks at different spent fuel basins followed by evaluation. (author)

  12. Development of experimental apparatus for evaluating corrosion resistance of cladding materials applied for advanced power reactor. 1

    International Nuclear Information System (INIS)

    Inohara, Yasuto; Ioka, Ikuo; Fukaya, Kiyoshi; Tachibana, Katsumi; Suzuki, Tomio; Kiuchi, Kiyoshi

    2001-03-01

    On the development of cladding materials for advanced power reactors, it is important to clarify long performance and to control the compatibility to high temperature water at heat conducting surfaces under heavy irradiation. On the present study, the high temperature water loop with an autoclave was made for examining the corrosion behavior up to the super critical water range and for developing the simulation testing technique under irradiation in the hot cell. The loop is applicable to immersion tests in the temperature and pressure ranges up to 450degC and 25 MPa that are covered the surface temperature range of fuel claddings. One of the characteristics of this apparatus is a pair of sapphire windows of autoclave for in-situ observations, and a phase transition from water to super critical water conditions was clearly verified through these windows. In this apparatus, it is possible to control the temperature, pressure and Dissolved Oxygen (DO) within a fluctuations of few % on three phases, namely, water, steam and super critical water. (author)

  13. Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

    Directory of Open Access Journals (Sweden)

    Simon Younan

    2018-01-01

    Full Text Available The objective of this study was to evaluate accident-tolerant fuel (ATF concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo and fully ceramic microencapsulated (FCM fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

  14. 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.

  15. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor

    International Nuclear Information System (INIS)

    Wintergerst, M.

    2008-01-01

    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  16. A finite element elastic-plastic analysis of residual stresses due to clad welding in reactor vessels

    International Nuclear Information System (INIS)

    Buchalet, C.; Riccardella, P.C.

    1972-01-01

    Residual stresses due to weld deposited cladding on the inside of a typical Westinghouse pressurized water reactor vessel are investigated using an axisymmetric finite element elastic-plastic analysis. At the beginning of the analysis, one head of the weld cladding is assumed to lie on the reactor vessel wall at melting temperature (2600degF), but in the solid phase, while the vessel remains at 300degF (preheat temperature). All material properties used in the calculations are taken as temperature-dependent. Temperature profiles are obtained in the cladding and base metal at several discrete time intervals. These temperatures profiles are used to obtain the stress distribution for the same time intervals. Residual hoop tensile stresses of approximately 25 ksi were found to exist in the cladding. Peak tensile stresses in the hoop direction occur in the base metal near the cladding interface and reach a value of 60 ksi at the end of the transient. The tensile stress decreases very rapidly through the thickness of the base metal and becomes insignificant at about two inches from the inside surface. In order to lower residual stresses, a post-weld heat treatment is performed by uniformly heating the vessel to 1100degF, holding at that temperature for a specified period of time and then cooling slowly. The analysis shows that after this treatment, the peak stresses in the base metal decrease from 60 ksi to 32 ksi, while the stress in the cladding does not change significantly. (author)

  17. Technological improvements to high temperature thermocouples for nuclear reactor applications

    International Nuclear Information System (INIS)

    Schley, R.; Leveque, J.P.

    1980-07-01

    The specific operating conditions of thermocouples in nuclear reactors have provided an incentive for further advances in high temperature thermocouple applications and performance. This work covers the manufacture and improvement of existing alloys, the technology of clad thermocouples, calibration drift during heat treatment, resistance to thermal shock and the compatibility of insulating materials with thermo-electric alloys. The results lead to specifying improved operating conditions for thermocouples in nuclear reactor media (pressurized water, sodium, uranium oxide) [fr

  18. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    International Nuclear Information System (INIS)

    Rebak, Raul B.

    2014-01-01

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  19. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  20. Thermo-mechanical assessment of full SiC/SiC composite cladding for LWR applications with sensitivity analysis

    Science.gov (United States)

    Singh, Gyanender; Terrani, Kurt; Katoh, Yutai

    2018-02-01

    SiC/SiC composites are considered among leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. Under a large heat flux application (i.e. fuel cladding), the non-uniform changes in the dimensions and physical properties will lead to build-up of stresses in the structure over the course of time. To ensure reliable and safe operation of such a structure it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and elevated temperature. In this work, the foundation for 3D thermo-mechanical analysis of SiC/SiC cladding is put in place and a set of analyses with simplified boundary conditions has been performed. The analyses were carried out with two different codes that were benchmarked against one another and prior results in the literature. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These robust models may now be modified to take into account the axial and circumferential variation in neutron and heat flux to fully account for 3D effects. The results from the simple analyses show the development of high tensile stresses especially in the circumferential and axial directions at the inner region of the cladding. Based on the results obtained, design guidelines are recommended. For lack of certainty in or tailor-ability for the physical and mechanical properties of SiC/SiC composite material a sensitivity analysis is conducted. The analysis results establish a precedence order of the properties based on the extent to which these properties influence the temperature and the stresses.

  1. Estimation of residual stresses in reactor pressure vessel steel specimens clad by stainless steel strip electrodes

    International Nuclear Information System (INIS)

    Schimmoeller, H.A.; Ruge, J.L.

    1978-01-01

    The equations to determine a two-dimensional state of residual stress in flat laminated plates are well known from an earlier work by one of the authors. The derivation of these equations leads to a linear, inhomogeneous system of Volterra's integral equations of the second kind. To ascertain the unknown residual stresses from these equations it is necessary to cut down the thickness of the test plate layer by layer. This results in two-dimensional deformation reactions in the rest of the test plate, which can be measured, e.g. by a strain gauge rosette applied to the opposite side of the plate. The above-mentioned stress analysis has been transferred to 86mm thick reactor pressure vessel steel specimens (Type 22NiMoCr 37, DIN-No. 1.6751, similar to ASTM A508, Class 2) double-run clad by austenitic stainless steel strip electrodes (first layer 24/13 Cr-Ni steel, second layer 21/10 Cr-Ni steel). The overall dimensions of the clad specimens investigated amounted to 200 x 200 x (86+4.5+4.5)mm. At the surface of the austenitic cladding there is a two-dimensional tensile normal stress state of about 200N/mm 2 parallel, and about 300N/mm 2 transverse, to the welding direction. The maximum tensile stress was 8mm below the interface (fusion line, material transition) in the parent material. The stress distributions of the specimens investigated, determined on the basis of the above-mentioned combined experimental mathematical procedure, are presented graphically for the as-welded (as-delivered) and annealed (600 0 C/12hr) conditions. (author)

  2. Eddy current examination of the nuclear fuel elements with aluminum 1100-F cladding of IPR-R1 research reactor: An initial study

    International Nuclear Information System (INIS)

    Silva, Roger F. da; Silva Júnior, Silvério F. da; Frade, Rangel T.; Rodrigues, Juliano S.

    2017-01-01

    Tubes of aluminum 1100-F as well as tubes of AISI 304 stainless steel are used as cladding of the fuel elements of TRIGA IPR-R1 nuclear research reactor. Usually, these tubes are inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements whose cladding has failed, but it is not able to determine the place where the discontinuity is located. On the other hand, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In previous works, the application of eddy current testing to evaluate the AISI 304 cladding fuel elements of TRIGA IPR-R1 was studied. In this paper, it is proposed an initial study about the use of eddy current testing for detection and characterization of discontinuities in the aluminum 1100-F fuel elements cladding. The study includes the development of probes and the design and manufacture of reference standards. (author)

  3. Eddy current examination of the nuclear fuel elements with aluminum 1100-F cladding of IPR-R1 research reactor: An initial study

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Roger F. da; Silva Júnior, Silvério F. da; Frade, Rangel T. [Centro de Desenvolvimento da Tecnologia Nucelar (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Rodrigues, Juliano S., E-mail: rfs@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: rtf@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Tubes of aluminum 1100-F as well as tubes of AISI 304 stainless steel are used as cladding of the fuel elements of TRIGA IPR-R1 nuclear research reactor. Usually, these tubes are inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements whose cladding has failed, but it is not able to determine the place where the discontinuity is located. On the other hand, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In previous works, the application of eddy current testing to evaluate the AISI 304 cladding fuel elements of TRIGA IPR-R1 was studied. In this paper, it is proposed an initial study about the use of eddy current testing for detection and characterization of discontinuities in the aluminum 1100-F fuel elements cladding. The study includes the development of probes and the design and manufacture of reference standards. (author)

  4. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Science.gov (United States)

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  5. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2008-01-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

  6. In-reactor fuel cladding external corrosion measurement process and results

    International Nuclear Information System (INIS)

    Thomazet, J.; Musante, Y.; Pigelet, J.

    1999-01-01

    Analysis of the zirconium alloy cladding behaviour calls for an on-site corrosion measurement device. In the 80's, a FISCHER probe was used and allowed oxide layer measurements to be taken along the outer generating lines of the peripheral fuel rods. In order to allow measurements on inner rods, a thin Eddy current probe called SABRE was developed by FRAMATOME. The SABRE is a blade equipped with two E.C coils is moved through the assembly rows. A spring allows the measurement coil to be clamped on each of the generating lines of the scanned rods. By inserting this blade on all four assembly faces, measurements can also be performed along several generating lines of the same rod. Standard rings are fitted on the device and allow on-line calibration for each measured row. Signal acquisition and processing are performed by LAGOS, a dedicated software program developed by FRAMATOME. The measurements are generally taken at the cycle outage, in the spent fuel pool. On average, data acquisition calls for one shift per assembly (eight hours): this corresponds to more than 2500 measurement points. These measurements are processed statistically by the utility program SAN REMO. All the results are collected in a database for subsequent behaviour analysis: examples of investigated parameters are the thermal/hydraulic conditions of the reactors, the irradiation history, the cladding material, the water chemistry This analysis can be made easier by comparing the behaviour measurement and prediction by means of the COROS-2 corrosion code. (author)

  7. The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Swam, L.F.P.; Shann, S.H.

    1991-01-01

    This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

  8. R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron Simon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tu, Lei [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize

  9. Reactor pressure vessel behaviour with a small crack in the cladding

    International Nuclear Information System (INIS)

    Fayolle, P.; Churier-Bossennec, H.; Faidy, C.

    1990-01-01

    This paper reports on fracture mechanic analysis of a PWR reactor pressure vessel with a 3.5 mm embedded circumferential crack in the cladding under a small lost of cooling accident transient. Different RTNDT level and effect of irradiation on material properties are considered. The study compares simplified one-dimensional and two-dimensional elastic approach and complete elastoplastic approach using J-parameter. The results show: good correlation between the different elastic approaches, important conservatism of the elastic approach compared to elastoplastic approach, no influence of irradiated material properties. The behavior of a vessel with this type of crack is acceptable for RTNDT less than 135 deg and safety injection temperature of 60 deg

  10. Study of reactions between nuclear fuel and cladding (316 stainless steel) in reactors. Influence of oxygen

    International Nuclear Information System (INIS)

    Otter, Monique.

    1980-12-01

    We have studied oxidation of 316 steel in close contact with oxides (Usub(0,74)Pusub(0,26)O 2 or UO 2 ), the stoichiometry of oxygen ranging from 2.00 to 2.5. Experiments are carried out either in a closed isothermal system or in an opened isothermal system with a fixed oxygen potential of uranium oxide. We have realized a potentiostatic device using a solid state electrotyte galvanic cell. In a closed system, the sensitized austenitic steel shows intergranular and volume oxidation probably enhanced by migration of steel components towards the fuel. Evidence of the usefulness of passivation have been obtained. We conclude that in a fast reactor sensitized cladding steel is oxydized by the constant potential of oxygen of UPuO 2 . Deposits observed in fuel can be explain by evaporation and cyclic transport phenomena that can be differents from VAN-ARKEL mechanism taking place through fission products [fr

  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. The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

    2006-01-01

    Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

  13. Sub-critical crack growth and clad integrity in a PWR reactor pressure vessel

    International Nuclear Information System (INIS)

    Tice, D.R.; Foreman, A.J.E.; Sharples, J.K.

    1987-10-01

    The possibility of in-service growth of sub-critical defects in a PWR reactor pressure vessel to a critical size which could result in vessel failure was addressed in both the 1976 and 1982 reports of the Light Water Reactor Study Group (LWRSG), under the Chairmanship of Dr W Marshall (now Lord Marshall). An addendum to this report was published by UKAEA in April 1987. The section of the addendum dealing with subcritical crack growth and the related issue of integrity of the stainless steel cladding on the inner vessel surface is reproduced in this report. This section of the LWRSG addendum provides a review of the current status of fatigue crack growth and environmentally assisted cracking research for pressure vessel steels in light water reactor environments, as well as a review of developments in crack growth assessment methods. The review concludes that the alternative assessment procedures now being developed give a more realistic prediction of in service crack growth than the ASME Section XI Appendix A fatigue crack growth curves. (author)

  14. 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.

  15. An application of the eddy-current method for inspections of austenitic cladding

    International Nuclear Information System (INIS)

    Kubis, S.; Herka, M.; Krajcovic, R.

    1999-01-01

    The application of the eddy-current method for inspections of austenitic cladding by means of pancake probes. Phase and amplitude characteristics of artificial defects and anticipated interference signals. Optimization of inspection parameters

  16. Study of cladding toughness in a pressure vessel steel water reactor

    International Nuclear Information System (INIS)

    Soulat, P.; Al Mundheri, M.

    1984-12-01

    Toughness of cladding and pressure vessel steel were determined at different temperatures in order to appreciate the participation of cladding resistance against crack propagation. The toughness of cladding is comparable with typical results on austenitic welds. The test on covered CT specimens shows the possibility of having a relatively good prevision of the behaviour of a coated structure

  17. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-01-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

  18. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-12-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

  19. Stress intensity factors for underclad and through clad defects in a reactor pressure vessel submitted to a pressurised thermal shock

    International Nuclear Information System (INIS)

    Marie, S.; Menager, Y.; Chapuliot, S.

    2005-01-01

    CEA has launched important work on the development of a Stress Intensity Factors compendium for cracks in a Reactor Pressure Vessel (RPV) taking into account the cladding. The work is performed by Finite Element analysis with a parametric mesh for two types of defects (under clad defect and through clad defect) and a wide range of geometrical and material parameters. In addition, an analytical stress solution for Pressurised Thermal Shock (PTS) on the RPV is proposed to allow a complete analytical estimation of the stress intensity factor K I for the PTS problem. The results are validated by comparison with a complete 3D finite element calculation performed on a complex and realistic case study

  20. Resistance welding of ODS cladding fuel a nuclear reactor of the fourth generation

    International Nuclear Information System (INIS)

    Corpace, F.

    2011-01-01

    ODS steels (Oxide Dispersion Strengthened) are candidate materials for fuel cladding in Sodium Fast Reactors (SFR), one of the studied concepts for the fourth generation of nuclear power plants. These materials possess good mechanical properties at high temperatures due to a dispersion of nano-meter-sized oxides into the matrix. Previous studies have shown that melting can induce a decrease in mechanical properties at high temperatures due to modifications of the nano-meter-sized oxide dispersion. Therefore the fusion welding techniques are not recommended and the solid state bonding has to be evaluated. This study is focused on resistance upset welding. Welding experiments and numerical simulations of the process are coupled in this thesis. All laboratory tests (experimental and numerical) are built using the experimental design method to evaluate the effects of the process parameters on the welding and on the weld. A 20Cr ODS steel is used for the experimental protocol. The first part is dedicated to the study of the influence of the process parameters on the welding process. The numerical simulations show that the welding steps can be divided in three stages. First, the contact temperature between the faying surfaces increases. The process is then driven in the second stage by the pieces geometry and especially the current constriction due to the thinness of the clad compared to the massive plug. Therefore, the heat generation is mainly located in the clad part out of the electrode leading to its collapse which is the third stage of the welding step. The evaluation of the process parameters influence on the physical phenomena (thermal, mechanical...) occurring during the welding step, allows adjusting them in order to influence thermal and mechanical solicitations undergone by the pieces during the welding process. The second part consists in studying the influence of physical phenomena on the welds. In the process parameter range, some welds exhibit compactness

  1. Plasma core reactor applications

    International Nuclear Information System (INIS)

    Latham, T.S.; Rodgers, R.J.

    1976-01-01

    Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

  2. Laser Cladding of Embedded Sensors for Thermal Barrier Coating Applications

    Directory of Open Access Journals (Sweden)

    Yanli Zhang

    2018-05-01

    Full Text Available The accurate real-time monitoring of surface or internal temperatures of thermal barrier coatings (TBCs in hostile environments presents significant benefits to the efficient and safe operation of gas turbines. A new method for fabricating high-temperature K-type thermocouple sensors on gas turbine engines using coaxial laser cladding technology has been developed. The deposition of the thermocouple sensors was optimized to provide minimal intrusive features to the TBC, which is beneficial for the operational reliability of the protective coatings. Notably, this avoids a melt pool on the TBC surface. Sensors were deposited onto standard yttria-stabilized zirconia (7–8 wt % YSZ coated substrates; subsequently, they were embedded with second YSZ layers by the Atmospheric Plasma Spray (APS process. Morphology of cladded thermocouples before and after embedding was optimized in terms of topography and internal homogeneity, respectively. The dimensions of the cladded thermocouple were in the order of 200 microns in thickness and width. The thermal and electrical response of the cladded thermocouple was tested before and after embedding in temperatures ranging from ambient to approximately 450 °C in a furnace. Seebeck coefficients of bared and embedded thermocouples were also calculated correspondingly, and the results were compared to that of a commercial standard K-type thermocouple, which demonstrates that laser cladding is a prospective technology for manufacturing microsensors on the surface of or even embedded into functional coatings.

  3. Corrosion of aluminium-clad spent fuel in LVR-15 research reactor storage facilities. Final report

    International Nuclear Information System (INIS)

    Splichal, K.; Berka, J.; Keilova, E.

    2006-03-01

    The corrosion of the research reactor aluminium clad spent fuel in water was investigated in two storage facilities. The standard racks were delivered by the IAEA and consisted of two aluminium alloys AA 6061 and Szav-1 coupons. Bimetallic couples create aluminium alloy and stainless steel 304 coupons. Rolled and extruded AA 6061 material was also tested. Single coupons, bimetallic and crevice couples were exposed in the at-reactor basin (ARB) and the high-level wastage pool (HLW). The water chemistry parameters were monitored and sedimentation of impurities was measured. The content of impurities of mainly Cl and SO 4 was in the range of 2 to 15 μg/l in the HLW pool; it was about one order higher in ARB. The Fe content was below 2 μg/l for both facilities. After two years of exposure the pitting was evaluated as local corrosion damage. The occurrence of pits was evaluated predominantly on the surfaces of single coupons and on the outer and inner surfaces of bimetallic and crevices coupons. No correlation was found between the pitting initiation and the type of aluminium alloys and rolled and extruded materials. In bimetallic couples the presence of stainless coupons did not have any effect on local corrosion. The depth of pits was lower than 50 μm for considerable areas of coupons and should be compared with the results of other participating institutes. (author)

  4. Influence of manufacturing process on the in-reactor creep anisotropy of stress-relieved Zircaloy-2 cladding

    International Nuclear Information System (INIS)

    Shann, S.H.; Van Swam, L.F.

    1995-01-01

    A procedure to determine the axial/radial and circumferential/radial contractile strain ratios (the R and P factors respectively in the Backofen-modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 cladding of BWR fuel rods, tie rods and water rods was developed and has been described previously (S.H. Shann and L.F. van Swam, Creep anisotropy of Zircaloy-2 cladding during irradiation, Trans. SMiRT-11, Vol. C, 1991). The present study employs the procedure to determine the anisotropy factors R and P for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding fabricated by various manufacturing processes. The analysis indicates that the cladding manufacturing process can have a pronounced effect on the anisotropy of irradiation-induced creep. Cladding types with identical yield and ultimate tensile strengths but fabricated by different manufacturing processes have different values of R and P during in-reactor creep. ((orig.))

  5. Thermoelastic analysis for the fuel claddings of the nuclear power reactor at Atucha in the skid's region

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, Gustavo; Basombrio, F.G.

    1979-01-01

    For the fuel elements of the Nuclear Power Reactor at Atucha, a two-dimensional thermoelastic analysis has been made in the region of the skids of the fuel cladding, when the gap between them and the fuel rod separator's support becomes zero. In such a case the latter forces exert on the skids an elastic reaction opposite to the cladding's expansion. The internal pressure reaching the yield stress for the cladding material has been calculated, as a function of the initial gap; for several possible fuel rod locations within the separator; for the actual dimensions and also for reduced thickness of the cladding; with a given external pressure and, with a known temperature spatial distribution. The latter has been calculated by solving the heat conduction equation along the fuel element for a certain power level in the reactor. The calculations are made with two FORTRAN IV computer codes developed at C.A.B., using the finite-element method: the NOLICUARM, to solve the nonlinear quasi-harmonic equation, and the ELASTEF 3, for the solution of thermoelastic problems with plane symmetry. (author) [es

  6. Materials development for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  7. UKAEA fast reactor project research and development programme on fuel element cladding and sub-assembly wrapper materials

    International Nuclear Information System (INIS)

    Harries, D.R.

    1977-01-01

    Research and development work on fuel element component (cladding, subassembly wrappers, etc.) materials for the U.K. sodium cooled fast reactor programme has been conducted at the United Kingdom Atomic Energy Authority (UKAEA) establishments at Dounreay, Harwell, Risley, and Springfields during the past fifteen years or so. This work has formed an integral part of, and has been co-ordinated by, the UKAEA Fast Reactor Project and has involved close liaison with the Nuclear Power Company (NPC) and the Central Electricity Generating Board (CEGB). The research and development were initially related to the Prototype Fast Reactor (PFR) but the scope has now been extended to cover the first Civil Fast Reactor (CFR1), which has recently been re-designated the Civil Demonstration Fast Reactor (CDFR). The paper outlines the present status of the development of sodium cooled fast reactors in the U.K. and proceeds to summarize the principal PFR and CDFR core and fuel element parameters which have determined the planning and direction of the fuel element materials programme. The current position on the fuel element cladding and wrapper research and development programme is reviewed, and the facilities and future irradiation programme to be carried out in PFR are described

  8. A regression model for zircaloy cladding in-reactor creepdown: Database, development, and assessment

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.; Lanning, D.

    1987-01-01

    The paper presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in a PWR and a BWR. This model accounts for variation in the zircaloy cladding heat treatments - cold worked and stress relieved material typically used in a PWR and fully recrystallized material typically used in a BWR. This model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. The paper also presents a comparison between cladding creep calculations by the creepdown model and corresponding test results from the KWU/CE program. ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the creepdown model calculates cladding creep strains reasonably well. (orig./HP)

  9. Application of laser cladding method to small-diameter stainless steel pipes in actual nuclear plant

    International Nuclear Information System (INIS)

    Atago, Y.; Yamadera, M.; Tsuji, H.; Shiraiwa, T.; Kanno, M.

    1995-01-01

    Recently, to prevent stress corrosion cracking (SCC) the material of stainless steel (Type 304), a laser cladding method which produces a highly corrosion-resisting coating (cladding) to be formed on the surface of the material was developed. This is applicable to a long distance and narrow space, 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 paste mixed metallic powder and heating resistive organic solvent is firstly placed on the inner surface of a small pipe and then a YAG laser beam transmitted through an optical fiber is irradiated to the paste, which will be melted and formed a clad subsequently, which is excellent in corrosion resistance. Finally, it can be achieved further resistance against the SCC due to the clad layer formed thus on the surface of the material. Recently, this Laser Cladding method was practically and successfully applied to the actual BWR Nuclear Power Plant in Japan. This report introduces the laser cladding technique, the equipments developed for practical application in the field

  10. Applications of Research Reactors

    International Nuclear Information System (INIS)

    2014-01-01

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

  11. 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

  12. 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.

  13. An example of coupling behaviour-damage-environment in polycrystals. Application to Pellet-Cladding Interaction

    International Nuclear Information System (INIS)

    Diard, Olivier

    2001-01-01

    Zircaloy-4 cladding is the first containment barrier for fission products, and its integrity must therefore be ensured in nominal and accidental situations. However, stress corrosion induced cracks may appear due to a strong pellet-cladding interaction. It is therefore important to model this interaction and crack growth and propagation to establish non-damage criteria. Thus, this research thesis aims at developing a modelling covering both issues (pellet-cladding interaction, and stress corrosion cracking) and allowing macroscopic and microscopic scales to be coupled. After a bibliographical synthesis on iodine-induced stress corrosion cracking and similar phenomena, the author presents the model proposed for the pellet-cladding interaction: phenomena to be taken into account, phenomenological and macroscopic behaviour laws used respectively for pellet and cladding. An extended version of an existing cladding viscoplastic model is proposed. Stress and strain fields in the cladding are obtained, notably in the contact zone. In the next part, the author presents various numerical tools developed or used to model multi-crystalline aggregates, and the model of crystalline plasticity used to simulate cladding behaviour at the microstructure scale. Effects of mesh density, element types and anisotropic elasticity are also discussed. The next chapter addresses the mechanical-chemical coupling. Some coupling formulas are presented for simple cases in order to define the effective diffusion coefficient. The last part reports the modelling of intergranular damage: definition of a damage criterion at the granular scale, assessment of stresses at grain boundaries, and effect of crystallographic neighbouring. A model of grain boundary damage is also proposed. This model is assessed on Failure Mechanics test samples and on simple microstructures. The application of the whole numerical model is reported [fr

  14. Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

    2014-08-15

    Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

  15. Behavior of underclad cracks in reactor pressure vessels - evaluation of mechanical analyses with tests on cladded mock-ups

    International Nuclear Information System (INIS)

    Moinereau, D.; Rousselier, G.; Bethmont, M.

    1993-01-01

    Innocuity of underclad flaws in the reactor pressure vessels must be demonstrated in the French safety analyses, particularly in the case of a severe transient at the end of the pressure vessel lifetime, because of the radiation embrittlement of the vessel material. Safety analyses are usually performed with elastic and elasto-plastic analyses taking into account the effect of the stainless steel cladding. EDF has started a program including experiments on large size cladded specimens and their interpretations. The purpose of this program is to evaluate the different methods of fracture analysis used in safety studies. Several specimens made of ferritic steel A508 C1 3 with stainless steel cladding, containing small artificial defects, are loaded in four-point bending. Experiments are performed at very low temperature to simulate radiation embrittlement and to obtain crack instability by cleavage fracture. Three tests have been performed on mock-ups containing a small underclad crack (with depth about 5 mn) and a fourth test has been performed on one mock-up with a larger crack (depth about 13 mn). In each case, crack instability occurred by cleavage fracture in the base metal, without crack arrest, at a temperature of about - 170 deg C. Each test is interpreted using linear elastic analysis and elastic-plastic analysis by two-dimensional finite element computations. The fracture are conservatively predicted: the stress intensity factors deduced from the computations (K cp or K j ) are always greater than the base metal toughness. The comparison between the elastic analyses (including two plasticity corrections) and the elastic-plastic analyses shows that the elastic analyses are often conservative. The beneficial effect of the cladding in the analyses is also shown : the analyses are too conservative if the cladding effects is not taken into account. (authors). 9 figs., 6 tabs., 10 refs

  16. State-of-the-art report on the development of liquid metal reactor fuel cladding materials in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Kuk, Il Hiun; Ryu, Woo Seog; Jang, Jin Sung; Rhee, Chang Kyu; Kim, Dae Whan; Park, Soon Dong; Kim, Woo Gon; Chung, Man Kyo; Han, Chang Hee

    1998-01-01

    PNC 1520 and PNC-FM5 have been developed as a cladding materials for LMR in Japan. PNC 1520 has superior swelling resistance and high temperature properties to PNC 31.6. And PNC-FMS steel has shown a high rupture stress as well as good neutron irradiation performance. In addition oxide dispersed ferritic steel (PNC-ODS) and 12Cr-8Mo steel have been developed. This report will give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is going to be operable in 2010 by analysis of the characteristics of cladding materials developed in Japan. (author). 39 refs., 2 tabs., 23 figs

  17. The fuel-cladding interfacial friction coefficient in water-cooled reactor fuel rods

    International Nuclear Information System (INIS)

    Smith, E.

    1979-01-01

    A central problem in the development of cladding failure criteria and of effective operational, design or material remedies is to know whether the cladding stress is enhanced significantly near cladding ridges, pellet chips or fuel pellet cracks; the latter may also be coincident with cladding ridges at pellet-pellet interfaces. As regards the fuel pellet crack source of cladding stress concentration, the magnitude of the uranium dioxide-Zircaloy interfacial friction coefficient μ governs the magnitude and distribution of the enhanced cladding stress. Considerable discussion, particularly at a Post-Conference Seminar associated with the SMIRT 4 Conference, has focussed on the value of μ, the author taking the view that it is unlikely to be large (< 0.5). The reasoning behind this view is as follows. A fuel pellet should fracture during a power ramp when the tensile hoop stress within the pellet exceeds the fuel's fracture stress. Since the preferred position for a fuel pellet crack to form is at the fuel-cladding interface midway between existing fuel cracks, where the interfacial shear stress changes sign, the pellet segment size after a power ramp provides a limit to the magnitude of the interfacial shear stresses and consequently to the value of μ. With this argument as a basis, the author's early work used the Gittus fuel rod model, in which there is a symmetric distribution of fuel pellet cracks and symmetric interfacial slippage, to show that μ < 0.5 if it is assumed that the average hoop stress within the cladding attains yield levels. It was therefore suggested that a high interfacial friction coefficient is unlikely to be operative during a power ramp; this result was used to support the view that interfacial friction effects do not play a dominant role in stress corrosion crack formation within the cladding. (orig.)

  18. 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.

  19. Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert W. [FPoliSolutions LLC, Murrysville, PA (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [FPoliSolutions LLC, Murrysville, PA (United States); Yurko, Joseph P. [FPoliSolutions LLC, Murrysville, PA (United States); Swindlehurst, Gregg [GS Nuclear Consulting, Charlotte, NC (United States); Zoino, Angelo [Univ. of Rome Tor Vergata (Italy)

    2015-09-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.

  20. Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

    International Nuclear Information System (INIS)

    Peacock, H.B. Jr.

    1999-01-01

    This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed

  1. Reduction in degree of absorber-cladding mechanical interaction by shroud tube in control rods for the fast reactor

    International Nuclear Information System (INIS)

    Donomae, Takako; Katsuyama, Kozo; Tachi, Yoshiaki; Maeda, Koji; Yamamoto, Masaya; Soga, Tomonori

    2011-01-01

    Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B 4 C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B 4 C pellets and the cladding tube. However, once B 4 C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B 4 C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B 4 C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported. (author)

  2. Preliminary experiments for the fabrication of clad for a spherical fuel for a research fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Almeida, L.A.A.

    1982-01-01

    A preliminary experiments using 1100 aluminum 0,5mm thick hemispheres welded on 10mm diameter steel and ceramic spheres in order to determine a method to clad spherical fuel for a research fluidized bed nuclear reactor were studied. The processes of hot press, T.I.G. and resistance we use for welding. A qualitative compression and metalographic tests of welded pieces are performed. By the analysis of the results of the tests we conclude that the resistance welding was the best. The experimental methods and the results with their analysis are presented in the paper. (Author) [pt

  3. FUMAC-a new model for light water reactor fuel relocation and pellet-cladding interaction

    International Nuclear Information System (INIS)

    Walton, L.A.; Matheson, J.E.

    1984-01-01

    An improved approach to the mechanical modeling of fuel rod performance is presented. Previous computer modeling has centered around a unified finite element approach with both fuel pellets and cladding being represented by ring elements. The fuel mechanical analysis code (FUMAC) departs from these approaches in two areas. The pellet model is an empirically based deterministic algorithm, while the cladding model uses both plane stress and plane strain finite elements. The work describes a semiempirical fuel cracking and fragment relocation model, which is burnup and power-level dependent. The interaction of the pellet with the cladding is treated classically. The resulting thick cylinder stresses are used in conjunction with an orthotropic creep model to predict cladding ridging. The resulting ridging compares well with experimental data for both steady-state and transient operating conditions. Future work planned includes the integration of the finite element cladding model with the pellet model and refinement of the pellet relocation and thermal models. Transient performance predictions will be emphasized

  4. Modelling of pellet-cladding interaction for PWRs reactors fuel rods

    International Nuclear Information System (INIS)

    Esteves, A.M.

    1991-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

  5. SIFAIL: a subprogram to calculate cladding deformation and damage for fast reactor fuel pins

    International Nuclear Information System (INIS)

    Wilson, D.R.; Dutt, D.S.

    1979-05-01

    SIFAIL is a series of subroutines used in conjunction with the thermal performance models of SIEX to assist in the evaluation of mechanical performance of mixed uranium plutonium oxide fuel pins. Cladding deformations due to swelling and creep are calculated. These have been compared to post-irradiation data from fuel pin tests in EBR-II. Several fuel pin cladding failure criteria (cumulative damage, total strain, and thermal creep strain) are evaluated to provide the fuel pin designer with a basis to select design parameters. SIFAIL allows the user many property options for cladding material. Code input is limited to geometric and environmental parameters, with a consistent set of material properties provided by the code. The simplified, yet adequate, thin wall stress--strain calculations provide a reliable estimate of fuel pin mechanical performance, while requiring a small amount of core storage and computer running time

  6. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Digby; Urquidi-Macdonald, Mirna; Chen, Yingzi; Ai, Jiahe; Park, Pilyeon; Kim, Han-Sang

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a

  7. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  8. A Multi-Layered Ceramic Composite for Impermeable Fuel Cladding for COmmercial Wate Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feinroth, Herbert

    2008-03-03

    A triplex nuclear fuel cladding is developed to further improve the passive safety of commercial nuclear plants, to increase the burnup and durablity of nuclear fuel, to improve the power density and economics of nuclear power, and to reduce the amount of spent fuel requiring disposal or recycle.

  9. 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.

  10. Advances in carbide fuel element development for fast reactor application

    International Nuclear Information System (INIS)

    Dienst, W.; Kleykamp, H.; Muehling, G.; Reiser, H.; Steiner, H.; Thuemmler, F.; Wedermeyer, H.; Weimar, P.

    1977-01-01

    The features of the carbide fuel development programme are reviewed and evaluated. Single pin and bundle irradiations are carried out under thermal, epithermal and fast flux conditions, the latter in the DFR and KNK-II reactors. Several fuel concepts in the region of representative SNR clad temperatures are compared by parameter and performance tests. A conservative concept is based on He-bonded 8 mm pins with (U,Pu)C pellets and a smear density of 75% TD, operating at 800 W/cm rod power and burnup to 70 MWd/kg. The preparation of mixed carbide fuels is carried out by carbothermic reduction of the oxides in different methods supported by equivalent carbon content, grain size and phase distribution analysis. The fuel for subassembly performance tests is produced in a pilot plant of 0,5 t/year capacity. Compatibility studies reveal that cladding carburization is the only chemical interaction with carbide fuels. This effect leads to a reduction in ductility of the stainless steel. Fission products apparently play no role in the compatibility behaviour. Comprehensive studies lead to reliable information on the chemical and thermodynamic state of the fuel under irradiation. The swelling of carbide fuels and the fission gas release are examined and analysed. Cladding plastic strain by fuel swelling occurs during steady-state operation because the irradiation creep is rather slow compared to oxide fuels. The cladding strain observed depends on the fuel porosity and the cladding strength. The development of carbide fuel pins is complemented by the application of comprehensive computer models. In addition to the steady-state tests power cycling and safety tests are under performance. Up to 1980 the results are summarized for the final design and specification. The development target of the present program is to fabricate several subassemblies for test operation in the SNR 300 by 1981

  11. Fabrication and inspection of stainless-steel-clad tubes for fast reactors

    International Nuclear Information System (INIS)

    Spriet, M.

    1975-01-01

    The production of cladding tubes requires a selection of the raw material, particular core taken during the cold and hot processes, special surface preparations, heat treatments, and intermediate control during the principal steps of fabrication. The inspection is made in two stages: acceptance tests at Vallourec (Eddy current and ultrasonic tests, metrology of internal and external diameter and thickness, metallography, analyses, tensile tests) and ultrasonic tests, metrology of external diameter and thickness, metallography, analyses, mechanical tests at high temperature) [fr

  12. J estimation scheme for cracks near the cladding of a reactor pressure vessel

    International Nuclear Information System (INIS)

    Fayolle, P.; Churier-Bossennec, H.; Faidy, C.

    1992-01-01

    The evaluation of flaws near the cladding is an important issue in term of risk of fast fracture of main vessel. This study analyses different K estimation schemes. These different K values are compared with respect to the toughness of the material K IC for different crack situations; the results confirm the validity of the proposal in the French RCC M Code for the plastic zone correction

  13. 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)

  14. Modelling the role of pellet crack motion in the (r-θ) plane upon pellet-clad interaction in advanced gas reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Haynes, T.A. [Centre for Nuclear Engineering & Department of Materials, Imperial College London, Exhibition Rd., London SW7 2AZ (United Kingdom); Ball, J.A. [EDF Energy, Barnett Way, Gloucester GL4 3RS (United Kingdom); Wenman, M.R., E-mail: m.wenman@imperial.ac.uk [Centre for Nuclear Engineering & Department of Materials, Imperial College London, Exhibition Rd., London SW7 2AZ (United Kingdom)

    2017-04-01

    Highlights: • Finite element modelling of pellet relocation in the (r-θ) plane of nuclear fuel. • ‘Soft’ and ‘hard’ PCI have been predicted in a cracked nuclear fuel pellet. • Stress concentration in the cladding ahead of radial pellet cracks is predicted. • The model is very sensitive to the coefficient of friction and power ramp duration. • The model is less sensitive to the number of cracks assumed. - Abstract: A finite element model of pellet fragment relocation in the r-θ plane of advanced gas-cooled reactor (AGR) fuel is presented under conditions of both ‘hard’ and ‘soft’ pellet-clad interaction. The model was able to predict the additional radial displacement of fuel fragments towards the cladding as well as the stress concentration on the inner surface resulting from the azimuthal motion of pellet fragments. The model was subjected to a severe ramp in power from both full power and after a period of reduced power operation; in the former, the maximum hoop stress in the cladding was found to be increased by a factor of 1.6 as a result of modelling the pellet fragment motion. The pellet-clad interaction was found to be relatively insensitive to the number of radial pellet crack. However, it was very sensitive to both the coefficient of friction used between the clad and pellet fragments and power ramp duration.

  15. Long term immersion test of aluminum alloy AA 6061 used for fuel cladding in MTR type reactors

    International Nuclear Information System (INIS)

    Linardi, Evelina M.; Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana

    2009-01-01

    In this work we present the results of long term immersion tests performed in the aluminum alloy AA 6061, used for fuel cladding in MTR type reactors. The tests were performed at open circuit potential in high purity water (ρ = 18.2 MΩ.cm) and in 10 -3 M NaCl solution. Two kinds of assemblies were studied: simple sheets and artificial crevices, immersed during 6, 12 and 18 months at room temperature. In both media and both assemblies, the aluminum hydroxide phases crystalline bayerite and bohemite were identified. It was found that a kind of localized attack named alkaline attack occurs around the iron-rich intermetallics. These particles were confirmed to control the corrosion of the AA 6061 alloy in an aerated medium. Immersion times for up to 18 months did not increase the oxide growth or the alkaline attack on the AA 6061 alloy. (author)

  16. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  17. In pile measurement of creep rate of stainless steel cladding tubes for fast reactor pins

    International Nuclear Information System (INIS)

    Calza Bini, A.; Cosoli, G.; Filacchioni, G.; Lanchi, M.; Nobili, A.; Pesce, E.; Rocca, U.V.; Rotoloni, P.L.

    1975-01-01

    Results are reported of a direct in pile measurement of creep on a cladding sample of 10cm length, under tensile stress of 22.82kg/mm 2 at a temperature of 550 0 during about 500 hours, up to an integrated flux of 2.6.10 20 n/cm 2 . Two identical samples were irradiated in the same temperature and flux conditions to be submitted to out of pile creep measurements together with other unirradiated samples. The aim of this first experiment was mainly to set up the device and to evaluate the kind and the quality of the available data

  18. Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Olshausen, K.D.

    1976-01-01

    A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

  19. Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Tverkovkin, B.E.

    1984-01-01

    The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

  20. 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

  1. Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

    International Nuclear Information System (INIS)

    Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

    1995-01-01

    The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

  2. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  3. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    International Nuclear Information System (INIS)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R.

    2015-01-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  4. An evaluation of the influence of fuel design parameters and burnup on pellet/cladding interaction for boiling water reactor fuel rod through in-core diameter measurement

    International Nuclear Information System (INIS)

    Yanagisawa, K.

    1986-01-01

    The influence of design parameters and burning on pellet/cladding interaction (PCI) of current boiling water reactor fuel rods was studied through in-core diameter measurement. Thinner cladding and a smaller diametral gap enhanced the PCI during startup. At constant power, fuel with SiO 2 added greatly reduced PCI due to relaxation. The fuel with a small grain size greatly reduced PCI due to densification. Preirradiation of rods up to 23 MWd/kgU caused a large PCI not only in a small gap but also in a large gap rod. Relaxation and permanent deformation was small. In the power increase experiment, one rod experienced PCI failure. The spurt times of coolant radioactivity coincided well with the sudden drop of cladding axial strain and marked crack opening at the rod surface. The estimated hoop stress predicted by FEMAXI-III was 350 MPa at the failure

  5. WWER reactor physics code applications

    International Nuclear Information System (INIS)

    Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.

    1994-01-01

    The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs

  6. Ultrasonic signal processing and B-SCAN imaging for nondestructive testing. Application to under - cladding - cracks

    International Nuclear Information System (INIS)

    Theron, G.

    1988-02-01

    Crack propagation under the stainless steel cladding of nuclear reactor vessels is monitored by ultrasonic testing. This work study signal processing to improve detection and sizing of defects. Two possibilities are examined: processing of each individual signal and simultaneous processing of all the signals giving a B-SCAN image. The bibliographic study of time-frequency methods shows that they are not suitable for pulses. Then decomposition in instantaneous frequency and envelope is used. Effect of interference of 2 close echoes on instantaneous frequency is studies. The deconvolution of B-SCAN images is obtained by the transducer field. A point-by-point deconvolution method, less noise sensitive, is developed. B-SCAN images are processed in 2 phases: interface signal processing and deconvolution. These calculations improve image accuracy and dynamics. Water-stell interface and ferritic-austenitic interface are separated. Echoes of crack top are visualized and crack-hole differentiation is improved [fr

  7. Study on fracture of fuel element cladding for naval reactor during typical accidents

    International Nuclear Information System (INIS)

    Zhang Fan; Shang Xueli; Zheng Zhongliang; Yu Lei

    2011-01-01

    Aiming at defining the grade of nuclear emergency response, the best estimate model has been adopted; the simulation of large break loss of coolant accident (LBLOCA) has been carried out by the radioactive analysis software coupled with relap5/mod 3.2 and core physics model. First, the peak clad temperature of the critical failure channel is calculated in relap5 code, and simultaneously its power factor is obtained. Second, pin power distribution of the fuel assemblies has been calculated in coarse-mesh nodal method. According to the pin power distribution in the whole core and the result gained above, the fraction of fuel element fracture is calculated. Finally, the radioactive analysis has been carried out and the reasonable source term is gotten, which can offer reference for the nuclear emergency decision making. (authors)

  8. Interaction between thorium and potential clad materials

    International Nuclear Information System (INIS)

    Kale, G.B.; Gawde, P.S.; Sengupta, Pranesh

    2005-01-01

    Thorium based fuels are being used for nuclear reactors. The structural stability of fuel-clad assemblies in reactor systems depend upon the nature of interdiffusion reaction between fuel-cladding materials. Interdiffusion reaction thorium and various cladding materials is presented in this paper. (author)

  9. Characterization and modeling of the thermal hydraulic and chemical environment of fuel claddings of PWR reactors during boiling

    International Nuclear Information System (INIS)

    March, Ph.

    1999-01-01

    In pressurised water reactors (PWR), nucleate boiling can strongly influence the oxidation rate of the fuel cladding. To improve our understanding of the effect of the boiling phenomenon on corrosion kinetics, information about the chemical and thermal hydraulic boundary conditions at the heating rod surface is needed. Moreover, very few data are available in the range of thermal hydraulic parameters of PWR cores (15,5 MPa and 340 deg C) concerning the two-phase flow pattern close to the fuel cladding. A visualization device has been adapted on an out-of-pile loop Reggae to obtain both qualitative and quantitative data. These observations provide a direct access to the geometrical properties of the vapor inclusions, the onset of nucleate boiling and the gas velocity and trajectory. An image processing method has been validated to measure both void fraction and interfacial area concentration in a bubbly two-phase flow. Thus, the visualization device proves to be a suitable and accurate instrumentation to characterize nucleate boiling in PWR conditions. The experimental results analysis indicates that a local approach is needed for the modelling of the fuel rod chemical environment. To simulate the chemical additives enrichment, a new model is proposed where the vapor bubbles are now considered as physical obstacles for the liquid access to the rod surface. The influence of the two-phase flow pattern appears to be of major importance for the enrichment phenomenon. This study clearly demonstrates the existence of strong interactions between the two-phase flow pattern, the rod surface condition, the corrosion process and the water chemistry. (author)

  10. Contribution to the replacement of cobalt-free hardfacing coating by laser cladding in fast neutron reactors

    International Nuclear Information System (INIS)

    De-Tran, Van

    2014-01-01

    This thesis contributes to the replacement of the coating of Stellite 6 which is used in friction areas for the primary circuit of the fast neutron reactor. It contains three parts: 1) A literature review for selecting the materials and the deposition process 2) A parametric study to get healthy deposits (good adhesion with the substrate, little porosity, no cracks, low dilution) 3) A study wear behavior of deposits obtained, at high temperature (200 C) under an atmosphere inert gas, to determine the wear resistance of materials selected without the influence of an eventual oxidation layer. From the literature review, it appears the following choices implemented in our study: * the method laser cladding with advantages such as: - Good adhesion (metallurgical) - High cooling speed - Low dilution rate - Wide parametric range * two nickel-based alloys: Colmonoy-52 and Tribaloy-700. These alloys have good dry wear behavior and could be deposited by the laser. In the manufacturing part of the healthy deposit, firstly, we characterized the metal powder. Then, a parametric study was performed to look for a good parametric range that makes us getting a healthy deposit of Stellite 6 (reference) of Colmonoy-52 and Tribaloy-700. In this case, relationships among three main process parameters laser cladding (laser beam power, surface scanning speed, rate of powder) with the microstructure and chemical composition of the deposit are studied. In study the wear behavior, a pin-on-disc type of tribological was used and tests were carried out in argon at room temperature and 200 C. We investigated the wear mechanism of the best deposition of Stellite 6, Colmonoy-52 and Tribaloy-700. The wear resistance of these materials were thoroughly compared. (author) [fr

  11. Hollow Core Photonic Crystal Fibre Comprising a Fibre Grating in the Cladding and its Applications

    DEFF Research Database (Denmark)

    2010-01-01

    An optical fibre is provided having a fibre cladding around a longitudinally extending optical propagation core. The cladding has a reflection region of a varying refractive index in the longitudinal direction....

  12. 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

  13. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  14. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  15. 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...

  16. Modeling of fast reactor cladding failure for hypothetical accident transient analysis

    International Nuclear Information System (INIS)

    Kramer, J.M.; DiMelfi, R.J.; Hughes, T.H.; Deitrich, L.W.

    1979-01-01

    An analysis is made of burst experiments performed on neutron irradiated cladding tubes. This is done by employing a generalized Voce equation to describe the mechanical deformation of type 316 stainless steel, combined with an empirical creep crack growth law, each modified to account for the effects of irradiation matrix hardening, and irradiation induced grain boundary embrittlement, respectively. The results of this analysis indicate that for large initial hoop stress, failure occurs at relatively low temperature and is controlled by the onset of plastic instability. The increase in failure temperature of irradiated material, in this low temperature region, is due to irradiation strengthening. Failure in the case of relatively small initial hoop stress occurs at high temperature where the Voce equation reduces to a power law creep formula. The ductility of irradiated material, in this high temperature region, is adequately described through the use of an empirical intergranular crack growth law used in conjunction with the creep law. The effect of neutron irradiation is to reduce the activation energy for crack propagation from the value for creep to some lower value correlated to independent Dorn rupture parameter measurements. The result is a predicted reduced ductility which translates into a reduction in failure temperature at a given hoop stress value for irradiated material. (orig.)

  17. Modeling deformation and failure of fast reactor cladding during simulated accident transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Dimelfi, R.J.

    1981-01-01

    An analysis is made of burst experiments performed on neutron irradiated cladding tubes. This is done by employing a generalized Voce equation to describe the mechanical deformation of type 316 stainless steel, combined with an empirical creep crack growth law, each modified to account for the effects of irradiation matrix hardening, and irradiation induced grain boundary embrittlement, respectively. The results of this analysis indicate that for large initial hoop stress, failure occurs at relatively low temperature and is controlled by the onset of plastic instability. The increase in failure temperature of irradiated material, in low temperature region, is due to irradiation strengthening. Failure in the case of relatively small initial hoop stress occurs at high temperature where the Voce equation reduces to a power law creep formula. The ductility of irradiated material, in this high temperature region, is adequately described through the use of an empirical intergranular crack growth law used in conjunction with the creep law. The effect of neutron irradiation is to reduce the activation energy for crack propagation from the value for creep to some lower value correlated to independent Dorn rupture parameter measurements. The result is a predicted reduced ductility which translates into a reduction in failure temperature at a given hoop stress value for irradiated material. (orig.)

  18. 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.)

  19. Corrosion properties of modified PNC1520 austenitic stainless steel in supercritical water as a fuel cladding candidate material for supercritical water reactor

    International Nuclear Information System (INIS)

    Nakazono, Yoshihisa; Iwai, Takeo; Abe, Hiroaki

    2009-01-01

    The supercritical water-cooled reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. There are some advantages including the use of a single phase coolant with high enthalpy. Supercritical Water (SCW) has never been used in nuclear power applications. There are numerous potential problems, particularly with materials. As the operating temperature of SCWR will be between 553 K and 893 K with a pressure of 25 MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel has been developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. Austenitic Fe-base steels were selected for possible use in supercritical water systems because of their corrosion resistance and radiation resistance. The PNC1520 austenitic stainless steel was selected for possible use in supercritical water systems. The corrosion data of PNC1520 in SCW is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in SCW. The SCW corrosion test was performed for the standard PNC1520 (1520S) and the Ti-additional type of PNC1520 (1520T) by using a SCW autoclave. The 1520S and 1520T are the first trial production materials of SCWR cladding candidate material in our group. Corrosion and compatibility tests on the austenitic 1520S and 1520T steels in supercritical water were performed at 673, 773 and 600degC with exposures up to 1000 h. We have evaluated the amount of weight gain, weight loss and weight of scale after the corrosion test in SCW for 1520S and 1520T austenitic steels. After 1000 h corrosion test performed, the weight gains of both austenitic stainless steels were less than 2 g/m 2 at 400degC and 500degC. But 1520T weight increases more and weight loss than 1520S at 600degC. The SEM observation result of the surface after 1000 h corrosion of an test

  20. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

    International Nuclear Information System (INIS)

    Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

    2007-01-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  1. Applications of the Dow TRIGA research reactor

    International Nuclear Information System (INIS)

    Kocher, C.W.; Quinn, T.J.; Krueger, D.A.

    1982-01-01

    The Dow TRIGA Mark I reactor is a one-hundred kilowatt nuclear reactor installed by General Atomics using the Torrey Pines reactor console, seventy-five used stainless-steel clad fuel elements and one new aluminium clad fuel element. The reactor is equipped with a forty-position rotating Lazy Susan in the reflector, a pneumatic transfer system with its terminal in the F-ring of the core, and a central thimble which can be used for irradiation of samples in the center of the core or which can be emptied of the shielding water to produce a beam of neutrons and gamma rays in the area atop the pool. Samples can also be irradiated in or near the core. There is no provision for pulsing this TRIGA reactor. The neutron activation analysis program uses the Dow TRIGA reactor as a source of thermal neutrons and a Kaman A711 generator as a source of 14-MeV neutrons. The associated counting equipment includes one Gel(Li) detector and two Nal(Tl) detectors, each using a 100-position sample changer and all interfaced to a Tracor-Northern TN-11 data acquisition and computing system, one Ge(Li) detector and its TN-11 system for the pneumatic transfer system and the beam tube experiments, and two NaKTl)detectors with a TN-4000 system used with the Kaman neutron generator. The activation analysis program gets samples from all parts of the manufacturing and research efforts at Dow: raw materials, intermediates, products, effluents, research samples, samples from customers who use Dow products, and environmental samples. This presentation is devoted to the progress made in the past year on the pneumatic transfer system and the renewed work on prompt gamma-ray spectroscopy including the extensive process of method validation

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  3. 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)

  4. Applicability of out-of-pile fretting wear tests to in-reactor fretting wear-induced failure time prediction

    Science.gov (United States)

    Kim, Kyu-Tae

    2013-02-01

    In order to investigate whether or not the grid-to-rod fretting wear-induced fuel failure will occur for newly developed spacer grid spring designs for the fuel lifetime, out-of-pile fretting wear tests with one or two fuel assemblies are to be performed. In this study, the out-of-pile fretting wear tests were performed in order to compare the potential for wear-induced fuel failure in two newly-developed, Korean PWR spacer grid designs. Lasting 20 days, the tests simulated maximum grid-to-rod gap conditions and the worst flow induced vibration effects that might take place over the fuel life time. The fuel rod perforation times calculated from the out-of-pile tests are greater than 1933 days for 2 μm oxidized fuel rods with a 100 μm grid-to-rod gap, whereas those estimated from in-reactor fretting wear failure database may be about in the range of between 60 and 100 days. This large discrepancy in fuel rod perforation may occur due to irradiation-induced cladding oxide microstructure changes on the one hand and a temperature gradient-induced hydrogen content profile across the cladding metal region on the other hand, which may accelerate brittleness in the grid-contacting cladding oxide and metal regions during the reactor operation. A three-phase grid-to-rod fretting wear model is proposed to simulate in-reactor fretting wear progress into the cladding, considering the microstructure changes of the cladding oxide and the hydrogen content profile across the cladding metal region combined with the temperature gradient. The out-of-pile tests cannot be directly applicable to the prediction of in-reactor fretting wear-induced cladding perforations but they can be used only for evaluating a relative wear resistance of one grid design against the other grid design.

  5. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    International Nuclear Information System (INIS)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-01-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  6. Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

    Science.gov (United States)

    Takeuchi, T.; Kameda, J.; Nagai, Y.; Toyama, T.; Matsukawa, Y.; Nishiyama, Y.; Onizawa, K.

    2012-06-01

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the δ-ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the δ-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the γ-austenite and δ-ferrite interface. There were no Cr depleted zones around the carbide.

  7. Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T.; Matsukawa, Y. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2012-06-15

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the {delta}-ferrite phase but not in the austenitic phase. Thermal aging at 400 Degree-Sign C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the {delta}-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the {gamma}-austenite and {delta}-ferrite interface. There were no Cr depleted zones around the carbide.

  8. Simulation of the cladding freezing during the loss of flow accident in a Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chu, N.N.; Eggen, D.T.

    1977-06-01

    The accident condition of the stainless steel cladding melting, relocation and freezing, following a loss of flow situation in gas-cooled fast reactors has not been determined yet. An alloy of 50 Pb/ 50 Sn was selected to facilitate experimental procedures because of its workability and its wide phase transition temperature range (183 to 216 0 C) similar to that of stainless steel (1375 to 1500 0 C). The objective of the experiment is to simulate the motion of liquid alloy through a tube and observe the conditions where it solidifies in the tube. The objective of the theoretical model is to use transient heat transfer analysis to describe the freezing front of the moving liquid metal and the plugging of the channel. Tests were conducted in alloy tubes having inside diameters 0.5 to 0.95 cm. Molten solder was poured through the vertically held tubes. The average falling velocity of the melt was measured to be about 89 cm/sec. The distance in the tube where the molten Pb/Sn solidifed across the diameter was measured. This penetration length varies from 20 to 40 cm as the initial liquid temperature ranges from 216 (liquidus point) to 500 0 C. The plugging time is calculated for those temperatures under which data on both the penetration distance and the falling velocity were obtained

  9. Computational geometry for reactor applications

    International Nuclear Information System (INIS)

    Brown, F.B.; Bischoff, F.G.

    1988-01-01

    Monte Carlo codes for simulating particle transport involve three basic computational sections: a geometry package for locating particles and computing distances to regional boundaries, a physics package for analyzing interactions between particles and problem materials, and an editing package for determining event statistics and overall results. This paper describes the computational geometry methods in RACER, a vectorized Monte Carlo code used for reactor physics analysis, so that comparisons may be made with techniques used in other codes. The principal applications for RACER are eigenvalue calculations and power distributions associated with reactor core physics analysis. Successive batches of neutrons are run until convergence and acceptable confidence intervals are obtained, with typical problems involving >10 6 histories. As such, the development of computational geometry methods has emphasized two basic needs: a flexible but compact geometric representation that permits accurate modeling of reactor core details and efficient geometric computation to permit very large numbers of histories to be run. The current geometric capabilities meet these needs effectively, supporting a variety of very large and demanding applications

  10. Development of underwater laser cladding and underwater laser seal welding techniques for reactor components

    International Nuclear Information System (INIS)

    Hino, Takehisa; Tamura, Masataka; Tanaka, Yoshimi; Kouno, Wataru; Makino, Yoshinobu; Kawano, Shohei; Matsunaga, Keiji

    2009-01-01

    Stress corrosion cracking (SCC) has been reported at the aged components in many nuclear power plants. Toshiba has been developing the underwater laser welding. This welding technique can be conducted without draining the water in the reactor vessel. It is beneficial for workers not to exposure the radiation. The welding speed can be attaining twice as fast as that of Gas Tungsten Arc Welding (GTAW). The susceptibility of SCC can also be lower than the Alloy 600 base metal. (author)

  11. Fracture mechanics analysis of reactor pressure vessel under pressurized thermal shock - The effect of elastic-plastic behavior and stainless steel cladding -

    International Nuclear Information System (INIS)

    Joo, Jae Hwang; Kang, Ki Ju; Jhung, Myung Jo

    2002-01-01

    Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). The PTS event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored

  12. 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

  13. Observations of in-reactor strain for fueled and unfueled FTR cladding

    International Nuclear Information System (INIS)

    Gilbert, E.R.; Makenas, B.J.; Wilson, D.R.

    1979-01-01

    It has been demonstrated that equations derived from in-reactor creep and swelling in unfueled pressurized tubes of 20% CW AISI 316 stainless steel can be used to predict strains in prototypic FTR mixed-oxide (UO 2 --PuO 2 ) fuel pins. For fast neutron fluences below 6 x 10 22 n/cm 2 the strains were small (less than one percent) and good agreement was found (within 0.1 percent diametral strain) between the equations and the fuel pin strains. This paper describes an extension of the earlier study to fast neutron fluences up to 11 x 10 22 n/cm 2

  14. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.

  15. Dynamics of a BWR with inclusion of boiling nonlinearity, clad temperature and void-dependent core power removal: Stability and bifurcation characteristics of advanced heavy water reactor (AHWR)

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2016-11-15

    Highlights: • Simplified models with inclusion of the clad temperature are considered. • Boiling nonlinearity and core power removal have been modeled. • Method of multiple time scales has been used for nonlinear analysis to get the nature and amplitude of oscillations. • Incorporation of modeling complexities enhances the stability of system. • We find that reactors with higher nominal power are more desirable from the point of view of global stability. - Abstract: We study the effect of including boiling nonlinearity, clad temperature and void-dependent power removal from the primary loop in the mathematical modeling of a boiling water reactor (BWR) on its dynamic characteristics. The advanced heavy water reactor (AHWR) is taken as a case study. Towards this end, we have analyzed two different simplified models with different handling of the clad temperature. Each of these models has the necessary modifications pertaining to boiling nonlinearity and power removal from the primary loop. These simplified models incorporate the neutronics and thermal–hydraulic coupling. The effect of successive changes in the modeling assumptions on the linear stability of the reactor has been studied and we find that incorporation of each of these complexities in the model increases the stable operating region of the reactor. Further, the method of multiple time scales (MMTS) is exploited to carry out the nonlinear analysis with a view to predict the bifurcation characteristics of the reactor. Both subcritical and supercritical Hopf bifurcations are present in each model depending on the choice of operating parameters. These analytical observations from MMTS have been verified against numerical simulations. A parametric study on the effect of changing the nominal reactor power on the regions in the parametric space of void coefficient of reactivity and fuel temperature coefficient of reactivity with sub- and super-critical Hopf bifurcations has been performed for all

  16. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    Hassan, Abobaker Mohammed Rahmtalla

    2014-09-01

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  17. Underwater cladding with laser beam and plasma arc welding

    International Nuclear Information System (INIS)

    White, R.A.; Fusaro, R.; Jones, M.G.; Solomon, H.D.; Milian-Rodriguez, R.R.

    1997-01-01

    Two welding processes, plasma arc (transferred arc) (PTA) and laser beam, were investigated to apply cladding to austenitic stainless steels and Inconel 600. These processes have long been used to apply cladding layers , but the novel feature being reported here is that these cladding layers were applied underwater, with a water pressure equivalent to 24 m (80 ft). Being able to apply the cladding underwater is very important for many applications, including the construction of off-shore oil platforms and the repair of nuclear reactors. In the latter case, being able to weld underwater eliminates the need for draining the reactor and removing the fuel. Welding underwater in reactors presents numerous challenges, but the ability to weld without having to drain the reactor and remove the fuel provides a huge cost savings. Welding underwater in reactors must be done remotely, but because of the radioactive corrosion products and neutron activation of the steels, remote welding would also be required even if the reactor is drained and the fuel removed. In fact, without the shielding of the water, the remote welding required if the reactor is drained might be even more difficult than that required with underwater welds. Furthermore, as shall be shown, the underwater welds that the authors have made were of high quality and exhibit compressive rather than tensile residual stresses

  18. Corrosion of research reactor aluminium-clad spent fuel in water-chemical and microbiological influenced

    International Nuclear Information System (INIS)

    Maksin, T.N.; Dobrijevic, R.P.; Idjakovic, Z.E.; Pesic, M.P.

    2002-01-01

    Spent fuel resulting from 25 years of operating research reactor RA at the Vinca Institute is presently all stored in the temporary spent fuel storage pool. It has been left in the ambient temperature and humidity for more then fifteen years so intensive corrosion processes were notice. We have spent fuel pools under control, after first research coordination meeting (RCM), of the first CRP, by monitoring of physical and chemical parameters of water in the pools, including temperature, pH-factor, electrical conductivity, mass concentration of corrosion products in the water and mud, mass concentration of relevant ions etc. The rack of standard corrosion coupons, was given at that time, has been in poor quality water for six years. We pick up rack assembly from basin and analysed. The results of this investigation are present in this article. (author)

  19. 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

  20. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection; Avaliacao de integridade de revestimentos de combustiveis de reatores de pesquisa e teste de materiais utilizando o ensaio de correntes parasitas

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete Anderson de

    2004-07-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  1. UK experience on fuel and cladding interaction in oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

    1977-04-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

  2. UK experience on fuel and cladding interaction in oxide fuels

    International Nuclear Information System (INIS)

    Batey, W.; Findlay, J.R.

    1977-01-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

  3. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  4. Irradiation experience with HT9-clad metallic fuel

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

    1991-01-01

    The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability

  5. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    2000-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  6. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1998-01-01

    Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  7. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1996-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  8. Improvement of the calculation of the stress intensity factors for underclad and through-clad defects in a reactor pressure vessel subjected to a pressurised thermal shock

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.

    2008-01-01

    The analysis of the stability of a defect in a cladded reactor pressure vessel (RPV) of a nuclear pressure water reactor (PWR) subjected to pressurised thermal shock (PTS) is one main elements of the general safety demonstration. Recently, CEA proposed several improved analytical tools for the analysis of the PTS. First, an analytical solution for the vessel through-thickness temperature variation has been developed to deal with any fluid temperature, taking into account the possible presence of a cladding, in the case of an internal PTS. The associated thermal stress expression has been simplified and a complete linearised solution is given for the thermal loading and also for internal pressure, depending on the main vessel material and on the cladding properties. Finally, a complete compendium is also given for the elastic stresses intensity factor calculation. This paper proposes several improvements of the proposed analytical method to deal with a PTS in a PWR cladded vessel. A variable heat transfer coefficient is now taken into account based on an equivalent fluid temperature variation determination, associated with a constant heat transfer coefficient, to keep the same thermal exchange between the fluid and the inner skin of the vessel obtained with the initial data. A more accurate expression for the linearised stresses due to the internal pressure is given, and a possible effect of residual stresses due to the difference between the operating temperature and the stress-free temperature is also taken into account. Finally, an extension of the domain of definition of the influence functions for the elastic stress intensity factor calculation is given

  9. Artificial intelligence applications to nuclear reactor diagnostics

    International Nuclear Information System (INIS)

    Lee, J.C.; Hassberger, J.A.; Wehe, D.K.

    1987-01-01

    The authors research into applications of artificial intelligence to nuclear reactor diagnostics involves three main areas. In the first area, the authors combine reactor simulation models and expert systems to diagnose the state of the plant. The second area examines ways in which the rule or knowledge base of an intelligent controller can be generated systematically from either fault trees or acquired plant data. Third, efforts are described to develop the capabilities to validate these techniques in a realistic reactor setting. The techniques are applicable to all reactor types, including fast reactors

  10. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  11. Annex 7 - Task 08/32 final report, Diffusion bond between the fuel and cladding of the RA reactor fuel element; Prilog 7 - Zavrsni izvestaj o podzadatku 08/32, Difuziona veza goriva i kosuljice gorivnog elementa reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Kustudic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Based on testing the diffusin bond between the fuel and the aluminium cladding of the RA reactor fuel element it was found that the diffusion bond exists; uranium is protected by a nickel layer to prevent the damaging diffusion between uranium and aluminium; the nickel layer is about 5 microns thick. Under assumption that the diffusion annealing was done after cladding the annealing temperature was about 500 deg C for about 1 hour.

  12. Activities performed within the program of nuclear safety research on structural and cladding materials for innovative reactor system able to transmute nuclear waste

    International Nuclear Information System (INIS)

    Fazio, C.; Rieth, M.; Lindau, R.; Aktaa, J.; Schneider, H-C.; Konys, J.; Yurechko, M.; Mueller, G.; Weisenburger, A.

    2009-01-01

    The transmutation of nuclear waste to reduce the burden on a geological repository is a relevant topic within the Program of Nuclear Safety Research of the Research Centre Karlsruhe. Several studies have confirmed that a high efficiency of transmutation of actinides is reached in fast neutron spectrum reactor system. Therefore, an important effort is dedicated to the study of transmutation strategies with different fast reactors and their associated technologies. Moreover, in international contexts as Generation IV International Forum (GIF) and Sustainable Nuclear Energy Technology Platform (SNETP), fast reactors are considered in the frame of sustainable development of nuclear energy and reduction of waste. The systems that are currently under investigation, in the frame of the different fuel cycle scenarios, are liquid metal cooled and gas cooled fast reactors as well as Accelerator Driven Sub-critical Transmutation devices (ADS). These innovative reactor systems, call for structural and clad materials, which are able to perform in a safe manner under the envisaged operational and postulated transient conditions. In this context the European Commission supports the FP7 project GETMAT, with the objective to contribute to the development and selection of reference structure materials for core components and primary systems of fast neutron reactors. Several institutes of the Research Centre Karlsruhe are involved in this project with activities in the area of 9Cr ODS steel development and mechanical characterisation; optimisation and ranking of weld and joining techniques as Electron Beam, TIG and Diffusion Bonding; assessment of materials behaviour in corrosive environment and in neutron and neutron/proton irradiation field; and development of corrosion protection barriers for cladding and primary system components and their characterisation. The objective of this contribution is to describe the context in which the GETMAT activities are embedded in the Program

  13. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Cupac, S.; Stanic, A.

    1990-01-01

    RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

  14. On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Holston, Anna-MariaAlvarez; Stjarnsater, Johan [Studsvik Nuclear AB, Nykoping (Sweden)

    2017-06-15

    Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (K{sub IH}) to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the K{sub IH}, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in K{sub IH} indicating the upper temperature limit for DHC. The value for K{sub IH} at 227°C was determined to be 2.6 ± 0.3 MPa √m.

  15. On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

    Directory of Open Access Journals (Sweden)

    Anna-Maria Alvarez Holston

    2017-06-01

    Full Text Available Delayed hydride cracking (DHC was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (KIH to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the KIH, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in KIH indicating the upper temperature limit for DHC. The value for KIH at 227°C was determined to be 2.6 ± 0.3 MPa √m.

  16. Development of new ferritic / martensitic steels for fuel cladding in fast neutron reactors

    International Nuclear Information System (INIS)

    Ratti, M.

    2009-11-01

    Many studies are directed toward the development of ferritic / martensitic ODS materials for applications in Gen IV programs. In this study, the mechanisms of formation of nano-phases (Y, Ti, O) and the influence of titanium on the precipitation refinement have been analyzed by small angle neutron scattering, X-ray diffraction and neutron diffraction. The obtained results allow developing new materials reinforced by nitrides (NDS which stands for Nitride Dispersion Strengthened). A first CEA patent is now being registered on these NDS materials processed by mechanical alloying. However, microstructural and mechanical characterizations are necessary to improve these new alloys. At last, a tensile and creep database has been acquired on an ODS Fe-18Cr material between room temperature and 650 C. These tests allow a qualitative description of the ODS mechanical behaviour. (author)

  17. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    D. Stahl

    2004-01-01

    The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO 2 , which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO 2 . The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes the

  18. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

  19. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  20. Effect of cladding defect size on the oxidation of irradiated spent LWR [light-water reactor] fuel below 3690C

    International Nuclear Information System (INIS)

    Einziger, R.E.; Strain, R.V.

    1984-01-01

    Tests on spent fuel fragments and rod segments were conducted between 250 and 360 0 C to relate temperature, defect size, and fuel oxidation rate with time-to-cladding-splitting. Defect sizes from 760 μm diameter down to 8 μm, the size of an SCC type breach, were used. Above 283 0 C, the time-to-cladding-splitting was longer for the smaller defects. The enhancement of the incubation time by smaller defects steadily decreased with temperature and was not detected at 250 0 C. 18 refs., 10 figs., 4 tabs

  1. Temperature and humidity effects on the corrosion of aluminium-base reactor fuel cladding materials during dry storage

    International Nuclear Information System (INIS)

    Peacock, H.B.; Sindelar, R.L.; Lam, P.S.

    2004-01-01

    The effect of temperature and relative humidity on the high temperature (up to 200 deg. C) corrosion of aluminum cladding alloys was investigated for dry storage of spent nuclear fuels. A dependency on alloy type and temperature was determined for saturated water vapor conditions. Models were developed to allow prediction of cladding behaviour of 1100, 5052, and 6061 aluminum alloys for up to 50+ years at 100% relative humidity. Calculations show that for a closed system, corrosion stops after all moisture and oxygen is used up during corrosion reactions with aluminum alloys. (author)

  2. Study of the aqueous corrosion mechanisms and kinetics of the AlFeNi aluminium based alloy used for the fuel cladding in the Jules Horowitz research reactor

    International Nuclear Information System (INIS)

    Wintergerst, M.

    2009-05-01

    For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)

  3. Two-Channel SPR Sensor Combined Application of Polymer- and Vitreous-Clad Optic Fibers.

    Science.gov (United States)

    Wei, Yong; Su, Yudong; Liu, Chunlan; Nie, Xiangfei; Liu, Zhihai; Zhang, Yu; Zhang, Yonghui

    2017-12-09

    By combining a polymer-clad optic fiber and a vitreous-clad optic fiber, we proposed and fabricated a novel optic fiber surface plasmon resonance (SPR) sensor to conduct two-channel sensing at the same detection area. The traditional optic fiber SPR sensor has many disadvantages; for example, removing the cladding requires corrosion, operating it is dangerous, adjusting the dynamic response range is hard, and producing different resonance wavelengths in the sensing area to realize a multi-channel measurement is difficult. Therefore, in this paper, we skillfully used bare fiber grinding technology and reverse symmetry welding technology to remove the cladding in a multi-mode fiber and expose the evanescent field. On the basis of investigating the effect of the grinding angle on the dynamic range change of the SPR resonance valley wavelength and sensitivity, we combined polymer-clad fiber and vitreous-clad fiber by a smart design structure to realize at a single point a two-channel measurement fiber SPR sensor. In this paper, we obtained a beautiful spectral curve from a multi-mode fiber two-channel SPR sensor. In the detection range of the refractive rate between 1.333 RIU and 1.385 RIU, the resonance valley wavelength of channel Ⅰ shifted from 622 nm to 724 nm with a mean average sensitivity of 1961 nm/RIU and the resonance valley wavelength of channel Ⅱ shifted from 741 nm to 976 nm with a mean average sensitivity of 4519 nm/RIU.

  4. Applications of computational intelligence in nuclear reactors

    International Nuclear Information System (INIS)

    Jayalal, M.L.; Jehadeesan, R.

    2016-01-01

    Computational intelligence techniques have been successfully employed in a wide range of applications which include the domains of medical, bioinformatics, electronics, communications and business. There has been progress in applying of computational intelligence in the nuclear reactor domain during the last two decades. The stringent nuclear safety regulations pertaining to reactor environment present challenges in the application of computational intelligence in various nuclear sub-systems. The applications of various methods of computational intelligence in the domain of nuclear reactors are discussed in this paper. (author)

  5. Application of the pulsed magnetic welding process to nuclear breeder reactor fuel pin end closures

    International Nuclear Information System (INIS)

    Brown, W.F.

    1984-01-01

    The pulsed magnetic welding process is a solid state welding process in which metallurgical bonding is effected by impacting metal or alloy parts against each other at high velocity by use of controlled high frequency, high intensity pulsed magnetic fields. This process is similar to the explosive welding process except that magnetic energy is used for impacting the parts together instead of using explosive energy. The pulsed magnetic welding (PMW) process is readily applied to the welding of cylindrical plugs to small diameter tubes. Although breeder reactor fuel pin design may vary in size, the application described here consisted of cladding tubes approximately 6.4 mm in diameter by 244 cm long with a wall thickness of 0.38 mm. After the cladding tubes are filled with fuel pellets and associated metal hardware, tapered end plugs are inserted into the end of the tubes and welded. A typical setup for PMW is described

  6. Two-Channel SPR Sensor Combined Application of Polymer- and Vitreous-Clad Optic Fibers

    Directory of Open Access Journals (Sweden)

    Yong Wei

    2017-12-01

    Full Text Available By combining a polymer-clad optic fiber and a vitreous-clad optic fiber, we proposed and fabricated a novel optic fiber surface plasmon resonance (SPR sensor to conduct two-channel sensing at the same detection area. The traditional optic fiber SPR sensor has many disadvantages; for example, removing the cladding requires corrosion, operating it is dangerous, adjusting the dynamic response range is hard, and producing different resonance wavelengths in the sensing area to realize a multi-channel measurement is difficult. Therefore, in this paper, we skillfully used bare fiber grinding technology and reverse symmetry welding technology to remove the cladding in a multi-mode fiber and expose the evanescent field. On the basis of investigating the effect of the grinding angle on the dynamic range change of the SPR resonance valley wavelength and sensitivity, we combined polymer-clad fiber and vitreous-clad fiber by a smart design structure to realize at a single point a two-channel measurement fiber SPR sensor. In this paper, we obtained a beautiful spectral curve from a multi-mode fiber two-channel SPR sensor. In the detection range of the refractive rate between 1.333 RIU and 1.385 RIU, the resonance valley wavelength of channel Ⅰ shifted from 622 nm to 724 nm with a mean average sensitivity of 1961 nm/RIU and the resonance valley wavelength of channel Ⅱ shifted from 741 nm to 976 nm with a mean average sensitivity of 4519 nm/RIU.

  7. 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)

  8. Inconel alloy 625 clad steel for application in wet scrubber systems

    International Nuclear Information System (INIS)

    Morse, S.L.; Shoemaker, L.E.

    1984-01-01

    Test panels from INCONEL 625 clad plate were successfully installed in two wet flue gas scrubber systems. In one system INCONEL 625 clad plate was located in the roof section of the absorber just ahead of the outlet ducting. The test plates, including weld seams, showed no signs to corrosion after six months of exposure. In the other scrubber test plates located in the outlet duct of an I.D. fan house, in the stack lining, and in the absorber quench area were unattacked after nine months

  9. Single-mode waveguides with SU-8 polymer core and cladding for MOEMS applications

    DEFF Research Database (Denmark)

    Nordström, Maria; Zauner, Dan; Boisen, Anja

    2007-01-01

    Fabrication and optical characterization of singlemode polymeric embedded waveguides are performed. A specific material combination (SU-8 2005 as core and the modified SU-8 mr-L 6050XP as cladding) is chosen in order to obtain a small refractive index difference for single-mode propagation combined...... can fabricate waveguides with an index difference in the order of 10−3, where both the core material and the cladding material are based on SU-8. The refractive index measurements are performed on thin polymeric films, while further optical characterizations are performed on waveguides with a height...

  10. A Comparative Study of Welded ODS Cladding materials for AFCI/GNEP Applications

    Energy Technology Data Exchange (ETDEWEB)

    Indrajit Charit; Megan Frary; Darryl Butt; K.L. Murty; Larry Zirker; James Cole; Mitchell Meyer; Rajiv S. Mishra; Mark Woltz

    2011-03-31

    This research project involved working on the pressure resistance welding of oxide dispersion strengthened (ODS) alloys which will have a large role to play in advanced nuclear reactors. The project also demonstrated the research collaboration between four universities and one nation laboratory (Idaho National Laboratory) with participation from an industry for developing for ODS alloys. These alloys contain a high number density of very fine oxide particles that can impart high temperature strength and radiation damage resistance suitable for in-core applications in advanced reactors. The conventional fusion welding techniques tend to produce porosity-laden microstructure in the weld region and lead to the agglomeration and non-uniform distribution of the neededoxide particles. That is why two solid state welding methods - pressure resistance welding (PRW) and friction stir welding (FSW) - were chosen to be evaluated in this project. The proposal is expected to support the development of Advanced Burner Reactors (ABR) under the GNEP program (now incorporated in Fuel Cycle R&D program). The outcomes of the concluded research include training of graduate and undergraduate students and get them interested in nuclear related research.

  11. A Comparative Study of Welded ODS Cladding materials for AFCI/GNEP Applications

    International Nuclear Information System (INIS)

    Charit, Indrajit; Frary, Megan; Butt, Darryl; Murty, K.L.; Zirker, Larry; Cole, James; Meyer, Mitchell; Mishra, Rajiv S.; Woltz, Mark

    2011-01-01

    This research project involved working on the pressure resistance welding of oxide dispersion strengthened (ODS) alloys which will have a large role to play in advanced nuclear reactors. The project also demonstrated the research collaboration between four universities and one nation laboratory (Idaho National Laboratory) with participation from an industry for developing for ODS alloys. These alloys contain a high number density of very fine oxide particles that can impart high temperature strength and radiation damage resistance suitable for in-core applications in advanced reactors. The conventional fusion welding techniques tend to produce porosity-laden microstructure in the weld region and lead to the agglomeration and non-uniform distribution of the needed oxide particles. That is why two solid state welding methods - pressure resistance welding (PRW) and friction stir welding (FSW) - were chosen to be evaluated in this project. The proposal is expected to support the development of Advanced Burner Reactors (ABR) under the GNEP program (now incorporated in Fuel Cycle R and D program). The outcomes of the concluded research include training of graduate and undergraduate students and get them interested in nuclear related research.

  12. 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

  13. Cladding nuclear steels - the application of plasma-arc hot wire surfacing

    International Nuclear Information System (INIS)

    Trarbach, K.O.

    1981-01-01

    The effect of one and two layer plasma-arc hot wire cladding on the HAZ microstructure of the fine grained structural steel 22 NiMoCr 3 7, which is similar to ASTM A 508, class 2, and steel 20 MnMoNi 5 5, similar to ASTM A 533, grade B, class 1 is determined. Attention is directed particularly to the behaviour of the susceptible region, and the consumables considered are cladding materials X 2 CrNiNb 19 9, similar to ER 347 Elc, and S-NiCr 20 Nb, similar to ER NiCr-3 (Inconel 82). Results of corrosion resistance tests show that this cladding technique can be recommended for manufacture of equipment for the chemical industry to avoid corrosion failure. Plasma-arc hot wire surfacing is also shown to be capable of depositing single or double clad layers to meet the highest safety requirements and could be applied to nuclear power plants for the special manufacture of wear resistant parts and for protection of equipment subject to a variety of corrosive environments. (U.K.)

  14. Clad Treatment in KARMA Code and Library

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-yeup; Lee, Hae-chan; Woo, Hae-seuk [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2016-05-15

    Zirconium is the main components in clad materials. The subgroup parameters of zirconium were generated with effective cross section which obtained by using flux distribution in clad region. It decreases absorption reaction rate differences with reference MCNP results. Use of composite nuclide is acceptable to increase efficiency but should be limited to specific target composition. Therefore, the use of the composite nuclide of Zircaloy-2 should be limited when HANA clad material is used for clad. Either using explicit components or generating composite nuclide for HANA is suggested. This paper investigates the clad analysis model for KARMA whether current method is applicable to HANA clad material.

  15. Application of Reactor Antineutrinos: Neutrinos for Peace

    Science.gov (United States)

    Suekane, F.

    2013-02-01

    In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (ν) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ν's carry direct and real time information useful for the safeguard purposes. Since ν can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

  16. Evaluation of Microstructural and Mechanical Property of Medium-sized HT9 Cladding Forged Material for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Lee, Kang Soo; Kim, Sung Ho; Lee, Chan Bock

    2012-01-01

    Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at 1170 degrees C after the induction melting to make round bar as 160 mm diameter, 7000 mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2-3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120 mm.

  17. Finite element application to global reactor analysis

    International Nuclear Information System (INIS)

    Schmidt, F.A.R.

    1981-01-01

    The Finite Element Method is described as a Coarse Mesh Method with general basis and trial functions. Various consequences concerning programming and application of Finite Element Methods in reactor physics are drawn. One of the conclusions is that the Finite Element Method is a valuable tool in solving global reactor analysis problems. However, problems which can be described by rectangular boxes still can be solved with special coarse mesh programs more efficiently. (orig.) [de

  18. Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

    Science.gov (United States)

    Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

    2018-02-01

    SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

  19. 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)

  20. Single-mode waveguides with SU-8 polymer core and cladding for MOEMS applications

    OpenAIRE

    Nordström, Maria; Zauner, Dan; Boisen, Anja; Hübner, Jörg

    2007-01-01

    Fabrication and optical characterization of singlemode polymeric embedded waveguides are performed. A specific material combination (SU-8 2005 as core and the modified SU-8 mr-L 6050XP as cladding) is chosen in order to obtain a small refractive index difference for single-mode propagation combined with the conventional fabrication method UV lithography to facilitate the integration of different types of optical detection methods on lab-on-a-chip systems. We analyze the behavior of the refrac...

  1. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  2. Low power reactor for remote applications

    International Nuclear Information System (INIS)

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs

  3. Laser Cladding of Ti-6Al-4V Alloy with Ti-Al2O3 Coating for Biomedical Applications

    Science.gov (United States)

    Mthisi, A.; Popoola, A. P. I.; Adebiyi, D. I.; Popoola, O. M.

    2018-05-01

    The indispensable properties of Ti-6Al-4V alloy coupled with poor tribological properties and delayed bioactivity make it a subject of interest to explore in biomedical application. A quite number of numerous coatings have been employed on titanium alloys, with aim to overcome the poor properties exhibited by this alloy. In this work, the possibility of laser cladding different ad-mixed powders (Ti - 5 wt.% Al2O3 and Ti - 8wt.% Al2O3) on Ti-6Al-4V at various laser scan speed (0.6 and 0.8 m/min) were investigated. The microstructure, phase constituents and corrosion of the resultant coatings were characterized by scanning electron microscope (SEM), Optical microscope, X-Ray diffractometer (XRD) and potentiostat respectively. The electrochemical behaviour of the produced coatings was studied in a simulated body fluid (Hanks solution). The microstructural results show that a defect free coating is achieved at low scan speed and ad-mixed of Ti-5 wt. % Al2O3. Cladding of Ti - Al2O3 improved the corrosion resistance of Ti-6Al-4V alloy regardless of varying neither scan speed nor ad-mixed percentage. However, Ti-5 wt.% Al2O3 coating produced at low scan speed revealed the highest corrosion resistance among the coatings due to better quality coating layer. Henceforth, this coating may be suitable for biomedical applications.

  4. Cladding properties under simulated fuel pin transients

    International Nuclear Information System (INIS)

    Hunter, C.W.; Johnson, G.D.

    1975-01-01

    A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

  5. In-fiber modal interferometer based on multimode and double cladding fiber segments for tunable fiber laser applications

    Science.gov (United States)

    Prieto-Cortés, P.; Álvarez-Tamayo, R. I.; Durán-Sánchez, M.; Castillo-Guzmán, A.; Salceda-Delgado, G.; Ibarra-Escamilla, B.; Kuzin, E. A.; Barcelata-Pinzón, A.; Selvas-Aguilar, R.

    2018-02-01

    We report an in-fiber structure based on the use of a multimode fiber segment and a double cladding fiber segment, and its application as spectral filter in an erbium-doped fiber laser for selection and tuning of the laser line wavelength. The output transmission of the proposed device exhibit spectrum modulation of the input signal with free spectral range of 21 nm and maximum visibility enhanced to more than 20 dB. The output spectrum of the in-fiber filter is wavelength displaced by bending application which allows a wavelength tuning of the generated laser line in a range of 12 nm. The use of the proposed in-fiber structure is demonstrated as a reliable, simple, and low-cost wavelength filter for tunable fiber lasers design and optical instrumentation applications.

  6. Insulator applications in a Tokamak reactor

    International Nuclear Information System (INIS)

    Leger, D.

    1986-06-01

    Insulators, among which insulators ceramics, have great potential applications in fusion reactors. They will be used for all plasma-facing components as protection and, magnetic fusion devices being subject to large electrical currents flowing in any parts of the device, for their electrical insulating properties

  7. Reactor applications of quantitative diffraction analysis

    International Nuclear Information System (INIS)

    Feguson, I.F.

    1976-09-01

    Current work in quantitative diffraction analysis was presented under the main headings of: thermal systems, fast reactor systems, SGHWR applications and irradiation damage. Preliminary results are included on a comparison of various new instrumental methods of boron analysis as well as preliminary new results on Zircaloy corrosion, and materials transfer in liquid sodium. (author)

  8. Ion thermometers - nuclear reactor applications

    International Nuclear Information System (INIS)

    Rosenkranz, J.; Jakes, D.

    The principle is briefly described of ion thermometers and the effects are reported of radiation on the ion crystal properties. The results show that ion thermometers are applicable for in-core measurements. (J.P.)

  9. Non-electric Applications of Fast Reactors

    International Nuclear Information System (INIS)

    Safa, Henri; Borgard, Jean-Marc

    2013-01-01

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

  10. 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)

  11. 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.

  12. Modelling of pellet-clad interaction during power ramps

    International Nuclear Information System (INIS)

    Zhou, G.; Lindback, J.E.; Schutte, H.C.; Jernkvist, L.O.; Massih, A.R.; Massih, A.R.

    2005-01-01

    A computational method to describe the pellet-clad interaction phenomenon is presented. The method accounts for the mechanical contact between fragmented pellets and the zircaloy clad, as well as for chemical reaction of fission products with zircaloy during power ramps. Possible pellet-clad contact states, soft, hard and friction, are taken into account in the computational algorithm. The clad is treated as an elastic-plastic-viscoplastic material with irradiation hardening. Iodine-induced stress corrosion cracking is described by using a fracture mechanics-based model for crack propagation. This integrated approach is used to evaluate two power ramp experiments made on boiling water reactor fuel rods in test reactors. The influence of the pellet-clad coefficient of friction on clad deformation is evaluated and discussed. Also, clad deformations, pellet-clad gap size and fission product gas release for one of the ramped rods are calculated and compared with measured data. (authors)

  13. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1975-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100 MWd/kg-heavy metal. The fuel is a sol-gel derived 88 atom-percent uranium (approximately 9 percent 235 U) 12 atom-percent plutonium oxide, and the cladding is 20 percent cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70 MWd/kg. The fuel has been operated at linear power rates of 39 and 44 kW/ m, and peak outer cladding temperature of 565 0 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48 kW/m (685 0 C). 4 references. (auth)

  14. Cladding creepdown model for FRAPCON-2

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.

    1985-02-01

    This report presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in both a pressurized water reactor (PWR) and a boiling water reactor (BWR). This model accounts for variations in zircaloy cladding heat treatment; cold worked and stress relieved material, typically used in a PWR, and fully recrystallized material, typically used in a BWR. The model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. This report also presents a comparison between cladding creep calculations by this model and corresponding measurements from the KWU/CE program, ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the model calculates cladding creep strains well. The analyses of non-fueled rods by FRAPCON-2 show that the cladding creepdown model was correctly incorporated. Also, analysis of a PWR rod test case shows that the FRAPCON-2 code can analyze pellet-cladding mechanical interaction caused by cladding creepdown and fuel swelling

  15. Friction Welding For Cladding Applications: Processing, Microstructure and Mechanical Properties of Inertia Friction Welds of Stainless Steel to Low Carbon Steel and Evaluation of Wrought and Welded Austenitic Stainless Steels for Cladding Applications in Acidchloride Service

    Science.gov (United States)

    Switzner, Nathan

    Friction welding, a solid-state joining method, is presented as a novel alternative process step for lining mild steel pipe and forged components internally with a corrosion resistant (CR) metal alloy for petrochemical applications. Currently, fusion welding is commonly used for stainless steel overlay cladding, but this method is costly, time-consuming, and can lead to disbonding in service due to a hard martensite layer that forms at the interface due to partial mixing at the interface between the stainless steel CR metal and the mild steel base. Firstly, the process parameter space was explored for inertia friction butt welding using AISI type 304L stainless steel and AISI 1018 steel to determine the microstructure and mechanical properties effects. A conceptual model for heat flux density versus radial location at the faying surface was developed with consideration for non-uniform pressure distribution due to frictional forces. An existing 1 D analytical model for longitudinal transient temperature distribution was modified for the dissimilar metals case and to account for material lost to the flash. Microstructural results from the experimental dissimilar friction welds of 304L stainless steel to 1018 steel were used to discuss model validity. Secondly, the microstructure and mechanical property implications were considered for replacing the current fusion weld cladding processes with friction welding. The nominal friction weld exhibited a smaller heat softened zone in the 1018 steel than the fusion cladding. As determined by longitudinal tensile tests across the bond line, the nominal friction weld had higher strength, but lower apparent ductility, than the fusion welds due to the geometric requirements for neck formation adjacent to a rigid interface. Martensite was identified at the dissimilar friction weld interface, but the thickness was smaller than that of the fusion welds, and the morphology was discontinuous due to formation by a mechanism of solid

  16. 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

  17. Clad-coolant chemical interaction

    International Nuclear Information System (INIS)

    Iglesias, F.C.; Lewis, B.J.; Desgranges, C.; Toffolon, C.

    2015-01-01

    This paper provides an overview of the kinetics for zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. Low-temperature oxidation of zircaloy due to water-side corrosion is further described. (authors)

  18. Erbium/ytterbium co-doped double clad fiber amplifier, its applications and effects in fiber optic communication systems

    Science.gov (United States)

    Dua, Puneit

    Increased demand for larger bandwidth and longer inter-amplifiers distances translates to higher power budgets for fiber optic communication systems in order to overcome large splitting losses and achieve acceptable signal-to-noise ratios. Due to their unique design ytterbium sensitized erbium doped, double clad fiber amplifiers; offer significant increase in the output powers that can be obtained. In this thesis we investigate, a one-stage, high power erbium and ytterbium co-doped double clad fiber amplifier (DCFA) with output power of 1.4W, designed and built in our lab. Experimental demonstration and numerical simulation techniques have been used to systematically study the applications of such an amplifier and the effects of incorporating it in various fiber optic communication systems. Amplitude modulated subcarrier multiplexed (AM-SCM) CATV distribution experiment has been performed to verify the feasibility of using this amplifier in an analog/digital communication system. The applications of the amplifier as a Fabry-Perot and ring fiber laser with an all-fiber cavity, a broadband supercontinuum source and for generation of high power, short pulses at 5GHz have been experimentally demonstrated. A variety of observable nonlinear effects occur due to the high intensity of the optical powers confined in micron-sized cores of the fibers, this thesis explores in detail some of these effects caused by using the high power Er/Yb double clad fiber amplifier. A fiber optic based analog/digital CATV system experiences composite second order (CSO) distortion due to the interaction between the gain tilt---the variation of gain with wavelength, of the doped fiber amplifier and the wavelength chirp of the directly modulated semiconductor laser. Gain tilt of the Er/Yb co-doped fiber amplifier has been experimentally measured and its contribution to the CSO of the system calculated. Theoretical analysis of a wavelength division multiplexed system with closely spaced

  19. Flaw behavior in mechanically loaded clad plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Robinson, G.C.; Oland, C.B.

    1989-01-01

    A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

  20. 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)

  1. FeCrAl/Zr dual layer fuel cladding for improved safety margin under accident scenario

    International Nuclear Information System (INIS)

    Park, D.J.; Park, J.H.; Jung, Y.I.; Kim, H.G.; Park, J.Y.; Koo, Y.H.

    2014-01-01

    For application of advanced steel as a cladding material in light water reactor (LWR), FeCrAl/Zr dual layer tube was manufactured by using a hot isostatic pressing (HIP) method. To optimize HIP condition for joining both FeCrAl and Zr alloys, HIP was carried out under various temperature conditions. Tensile test and 3-point bend test performed for measuring mechanical properties of HIPed sample. To better understand microstructural characteristics in interface region between two alloys, SEM and TEM study were conducted by using HIPed sample with different process conditions. Based on this optimization study and analyzed results, optimized HIP condition was determined and FeCrAl/Zr dual layer fuel cladding having same wall thickness with current LWR fuel cladding was manufactured. Simulated loss-of-coolant accident test was carried out using FeCrAl/Zr dual layer cladding sample and fuel integrity was measured by mechanical test. (authors)

  2. Compatibility studies on Mo-coating systems for nuclear fuel cladding applications

    Science.gov (United States)

    Koh, Huan Chin; Hosemann, Peter; Glaeser, Andreas M.; Cionea, Cristian

    2017-12-01

    To improve the safety factor of nuclear power plants in accident scenarios, molybdenum (Mo), with its high-temperature strength, is proposed as a potential fuel-cladding candidate. However, Mo undergoes rapid oxidation and sublimation at elevated temperatures in oxygen-rich environments. Thus, it is necessary to coat Mo with a protective layer. The diffusional interactions in two systems, namely, Zircaloy-2 (Zr2) on a Mo tube, and iron-chromium-aluminum (FeCrAl) on a Mo rod, were studied by aging coated Mo substrates in high vacuum at temperatures ranging from 650 °C to 1000° for 1000 h. The specimens were characterized using scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS) and nanoindentation. In both systems, pores in the coating increased in size and number with increasing temperature over time, and cracks were also observed; intermetallic phases formed between the Mo and its coatings.

  3. Application of Copper Cladding Aluminum Composites in UHV Portable Earthing and Short-circuiting Wires

    Directory of Open Access Journals (Sweden)

    Zhu Jianjun

    2018-01-01

    Full Text Available Aiming at the heavy weight and inconvenience when carrying and installing copper earthing wires on the UHV transmission lines, in this paper, we present the use of copper clad aluminum(CCA composite materials as a lightweight method for UHV earthing wire conductor. Theoretical calculations and tests of the fusing current in a short time for copper and CCA material are conducted. The results show that the theoretical value of the earthing wire conductor's fusing current corresponds with the test value on condition of the conductor cross section greater than 4mm2 as well as fusing time less than 1.5s. The CCA-10 earthing wires get 36.2% weight reduction compared with copper wires.

  4. TORFA - toroidal reactor for fusion applications

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1980-09-01

    The near-term goal of the US controlled fusion program should be the development, for practical applications, of an intense, quasi-steady, reliable 14-MeV neutron source with an electrical utilization efficiency at least 10 times larger than the value characterizing beam/solid-target neutron generators. This report outlines a method for implementing that goal, based on tokamak fusion reactors featuring resistive toroidal-field coils designed for ease of demountability

  5. Advances in zirconium technology for nuclear reactor application

    International Nuclear Information System (INIS)

    Ganguly, C.

    2002-01-01

    Zirconium alloys are extensively used as a material for cladding nuclear fuels and for making core structurals of water-cooled nuclear power reactors all over the world for generation of nearly 16 percent of the worlds electricity. Only four countries in the world, namely France, USA, Russia and India, have large zirconium industry and capability to manufacture reactor grade zirconium sponge, a number of zirconium alloys and a wide variety of structural components for water cooled nuclear reactor. The present paper summarises the status of zirconium technology and highlights the achievement of Nuclear Fuel Complex during the last ten years in developing a wide variety of zirconium alloys and components for water-cooled nuclear power programme

  6. Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

    1982-05-01

    Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

  7. 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.

  8. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barani, Tommaso [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pizzocri, Davide [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  9. In reactor performance of defected zircaloy-clad U{sub 3}Si fuel elements in pressurized and boiling water coolants

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Allison, G M; Ambler, J F.R.; Chalder, G H; Lipsett, J J

    1968-05-15

    The results of two in-reactor defect tests of Zircaloy-clad U{sub 3}Si are reported. In the first test, a previously irradiated element ({approx}5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at {approx}270{sup o}C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U{sub 3}Si at 300{sup o}C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  10. 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.

  11. Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    International Nuclear Information System (INIS)

    Batte, G.L.; Hoffman, G.L.

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

  12. Manufacturing of FeCrAl/Zr Dual Layer tube for its application to LWR Fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    Many advanced materials such as MAX phases, Mo, SiC, and Fe-based alloys are being considered a possible candidate to substitute the Zr-based alloy cladding has been used in light water reactors. Among the proposed candidate materials, Fe-based alloy is one of the most promising candidates owing to its excellent formability, very good high strength, and corrosion resistance at high temperature. However, neutron cross section of FeCrAl alloy is much higher than that of existing Zr-based alloys. In this study, FeCrAl/Zr dual layer tube was manufactured by using a hot isostatic pressing (HIP) method. The thickness of outer FeCrAl layer was varied from 50 to 250 μm but all the FeCrAl/Zr dual layer tube samples maintained its total thickness of 570 μm. For a detailed microstructural characterization of FeCrAl/Zr dual layer, polarized optical microscopy and scanning electron microscopy (SEM) study carried out and its mechanical property was measured by ring compression test. FeCrAl/Zr dual layer tube sample was successfully manufactured with good adhesion between both layers. Inter layer showing gradual element variation was observed at interface. Result obtained from simulated LOCA test indicates that FeCrAl/Zr dual layer tube may maintain its integrity during LOCA and its accident tolerance had greatly improved compared to that of Zr-based alloy.

  13. 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.

  14. Automatic mesh refinement and local multigrid methods for contact problems: application to the Pellet-Cladding mechanical Interaction

    International Nuclear Information System (INIS)

    Liu, Hao

    2016-01-01

    This Ph.D. work takes place within the framework of studies on Pellet-Cladding mechanical Interaction (PCI) which occurs in the fuel rods of pressurized water reactor. This manuscript focuses on automatic mesh refinement to simulate more accurately this phenomena while maintaining acceptable computational time and memory space for industrial calculations. An automatic mesh refinement strategy based on the combination of the Local Defect Correction multigrid method (LDC) with the Zienkiewicz and Zhu a posteriori error estimator is proposed. The estimated error is used to detect the zones to be refined, where the local sub-grids of the LDC method are generated. Several stopping criteria are studied to end the refinement process when the solution is accurate enough or when the refinement does not improve the global solution accuracy anymore. Numerical results for elastic 2D test cases with pressure discontinuity show the efficiency of the proposed strategy. The automatic mesh refinement in case of unilateral contact problems is then considered. The strategy previously introduced can be easily adapted to the multi-body refinement by estimating solution error on each body separately. Post-processing is often necessary to ensure the conformity of the refined areas regarding the contact boundaries. A variety of numerical experiments with elastic contact (with or without friction, with or without an initial gap) confirms the efficiency and adaptability of the proposed strategy. (author) [fr

  15. 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.

  16. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    International Nuclear Information System (INIS)

    Olsen, A.R.; Judkins, R.R.

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O 2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

  17. Irradiation temperature memorization by retention of krypton-85. Application to the temperature determination for the internal cladding surface of fuel elements in PWR

    International Nuclear Information System (INIS)

    Fremiot, Claude

    1977-01-01

    The temperature of the inner surface of the cladding fuel elements, which can not be measured directly, can be determined after irradiation. During its stage within the reactor, the cladding is bombarded by krypton-85 fission product, which is trapped in the metallic lattice defects. The experience shows that the krypton release during postirradiation heating takes place at the irradiation temperature. This method was applied for PWR fuel element. A very simple model for retention and release of the krypton is proposed. The krypton trap-energy in zircaloy partakes in this model. This technique can be ordered amongst the Hot'Lab' control methods and expert appraisements. It is pointed out that the principal interest in that method is the fact that it does not need any fuel element instrumentation. At the present, this method is being used by CEA for routine-control. [fr

  18. Tasks related to increase of RA reactor exploitation and experimental potential, 02. Verification of the system for detecting failures of the RA reactor fuel element cladding; Radovi na povecanju eksploatacionih i eksperimentalnih mogucnosti reaktora RA, 02. Provera sistema za detekciju pucanja kosuljice gorivnog elementa reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    For the purpose of this task it was necessary to analyze the time dependent distribution of fission products in the fuel element and leaks through the cracks in the cladding; to calculate the quantity of solid fission products, volatile fission products and fission gases for the RA reactor; to collect the data for estimating the activity of the short-living isotopes created by neutron irradiation of D{sub 2}O; analyze the number of delayed neutrons in D{sub 2}O. Experiment were needed to estimate the distribution of Xe and Kr in the heavy water in the reactor channel and analyze the activity of D{sub 2}O and helium based on reactor operation data. For the purpose of verifying the efficiency and safety of the existing system for detecting the cracks of the fuel element cladding is presented in this report together with the review of similar systems at a number of reactors in the world.

  19. R-matrix parameters in reactor applications

    International Nuclear Information System (INIS)

    Hwang, R.N.

    1992-01-01

    The key role of the resonance phenomena in reactor applications manifests through the self-shielding effect. The basic issue involves the application of the microscopic cross sections in the macroscopic reactor lattices consisting of many nuclides that exhibit resonance behavior. To preserve the fidelity of such a effect requires the accurate calculations of the cross sections and the neutron flux in great detail. This clearly not possible without viable resonance data. Recently released ENDF/B VI resonance data in the resolved range especially reflect the dramatic improvement in two important areas; namely, the significant extension of the resolved resonance ranges accompanied by the availability of the R-matrix parameters of the Reich-Moore type. Aside from the obvious increase in computing time required for the significantly greater number of resonances, the main concern is the compatibility of the Riech-Moore representation to the existing reactor processing codes which, until now, are based on the traditional cross section formalisms. This purpose of this paper is to summarize our recent efforts to facilitate implementation of the proposed methods into the production codes at ANL

  20. Review on experience in operation of steels for use as fast reactor fuel element cladding and wrappers

    International Nuclear Information System (INIS)

    Weisz, Michel.

    1978-01-01

    The informations on the behavior of steels which can be gathered from routine destructive or non-destructive examination of fast reactor cans and wrapper tubes are presented. The relative merits of swelling measurements made on specimens or on real cans are compared. The diametral deformations of all the cans of a bundle and the immersion density measurements on industrial fabrication of wrapper tubes are discussed. The swelling temperature relationships and the double peak swelling of SA316 in relation with microstructural evolution are studied. Irradiation creep is also investigated, particularly the bulging of the wrapper tubes allowing to derive mean creep rate measurements for the high dose region [fr

  1. Prospects for applications of ship-propulsion nuclear reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.

    1994-01-01

    The use of ship-propulsion nuclear power reactors in remote areas of Russia is examined. Two ship reactors were analyzed: the KLT-40, a 170 MW-thermal reactor; and the KN-3, a 300 MW-thermal reactor. The applications considered were electricity generation, desalination, and drinking water production. Analyses showed that the applications are technically justified and could be economically advantageous. 5 refs., 9 figs., 1 tab

  2. Thermal stress intensity factor for an axial crack in a clad cylinder

    International Nuclear Information System (INIS)

    Kuo, An Yu; Deardorf, A.F.; Riccardella, P.C.

    1993-01-01

    Many clad pressure vessels have been found to have cracks running through the inside surface cladding and into the base material. Although Young's moduli and Poisson's ratios of the clad and base materials are about the same for most of the industrial applications, coefficients of thermal expansion of the two dissimilar materials, clad and base materials, are usually quite different. For example, low alloy ferritic steel is a common base material for reactor pressure vessels (RPV) and the vessels are usually clad with austenitic stainless steel. Young's moduli for the low alloy steel and stainless steel at 350 F are 29,000 ksi and 28,000 ksi, respectively, while their coefficients of thermal expansion are 7.47x10 -6 in/in and 9.50x10 -6 in/in-degree F, respectively. The mismatch in coefficients of thermal expansion will cause high residual thermal stress even when the entire vessel is at a uniform temperature. This residual stress is one of the primary reasons why so many cracks have been found in the cladded components. In performing reactor pressure vessel integrity evaluation, such as computing probability of brittle fracture of the RPV, it is necessary to calculate stress intensity factors for cracks, which initiate from the clad material and run into the base metal. This paper presents a convenient method of calculating stress intensity factor for an axial crack emanating from the inside surface of a cladded cylinder under thermal loading. A J-integral like line integral was derived and used to calculate the stress intensity factors from finite element stress solutions of the problem

  3. Reactor alarm system development and application issues

    Energy Technology Data Exchange (ETDEWEB)

    Drexler, J E; Oicese, G O [INVAP S.E. (Argentina)

    1997-09-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ``Reactor Alarm System`` (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: (a) Alarm Panel Display; (b) Alarm Page; (c) Alarm Masking and Inhibition; (d) Alarms Color and Attributes; (e) Condition Classification; and (f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: (a) Multiple-copy alarm summaries; (b) Improved alarm handling; (c) Extended dictionary; and (d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs.

  4. Reactor alarm system development and application issues

    International Nuclear Information System (INIS)

    Drexler, J.E.; Oicese, G.O.

    1997-01-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

  5. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional {delta}-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the {epsilon}/d model, was developed. The correct solution has been derived in the {epsilon}/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In

  6. Development and application of reactor noise diagnostics

    International Nuclear Information System (INIS)

    Karlsson, Joakim K.H.

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional δ-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the ε/d model, was developed. The correct solution has been derived in the ε/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In the paper

  7. Vanadium alloys for fusion reactor applications

    International Nuclear Information System (INIS)

    Mattas, R.F.; Loomis, B.A.; Smith, D.L.

    1992-01-01

    This paper reports that fusion reactors will produce a severe operating environment for structural materials. The material should have good mechanical strength and ductility to high temperature, be corrosion resistant to the local environment, have attractive thermophysical properties to accommodate high heat loads, and be resistant to neutron damage. Vanadium alloys are being developed for such applications, and they exhibit desirable properties in many areas Recent progress in vanadium alloy development indicates good strength and ductility to 700 degrees C, minimal degradation by neutron irradiation, and reduced radioactivity compared with other candidate alloy systems

  8. Evolution of the thickness of the aluminum oxide film due to the pH of the cooling water and surface temperature of the fuel elements clad of a nuclear reactor

    International Nuclear Information System (INIS)

    Babiche, Ivan

    2013-01-01

    This paper describes the mechanism of growth of a film of aluminum oxide on an alloy of the same material, which serves as a protective surface being the constituent material of the RP-10 nuclear reactor fuel elements clads. The most influential parameters on the growth of this film are: the pH of the cooling water and the clad surface temperature of the fuel element. For this study, a mathematical model relating the evolution of the aluminum oxide layer thickness over the time, according to the same oxide film using a power law is used. It is concluded that the time of irradiation, the heat flux at the surface of the aluminum material, the speed of the coolant, the thermal conductivity of the oxide, the initial thickness of the oxide layer and the solubility of the protective oxide are parameters affecting in the rate and film formation. (author).

  9. Some proposed mechanisms for internal cladding corrosion

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Whitlow, W.H.

    1977-01-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  10. Some proposed mechanisms for internal cladding corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M H; Pickering, S; Whitlow, W H [EURATOM (United Kingdom)

    1977-04-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  11. LASER SURFACE CLADDING FOR STRUCTURAL REPAIR

    OpenAIRE

    SANTANU PAUL

    2018-01-01

    Laser cladding is a powder deposition technique, which is used to deposit layers of clad material on a substrate to improve its surface properties. It has widespread application in the repair of dies and molds used in the automobile industry. These molds and dies are subjected to cyclic thermo-mechanical loading and therefore undergo localized damage and wear. The final clad quality and integrity is influenced by various physical phenomena, namely, melt pool morphology, microst...

  12. Pseudo-harmonics method: an application to thermal reactors

    International Nuclear Information System (INIS)

    Silva, F.C. da; Rotenberg, S.; Thome Filho, Z.D.

    1985-10-01

    Several applications of the Pseudo-Harmonics method are presented, aiming to calculate the neutron flux and the perturbed eigenvalue of a nuclear reactor, like PWR, with three enrichment regions as Angra-1 reactor. In the reference reactor, perturbations of several types as global as local were simulated. The results were compared with those from the direct calculation. (E.G.) [pt

  13. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    Ponsard, B.

    2005-01-01

    The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

  14. Application of laser cladding to the aeroengine component. Koku engine buhin eno laser nikumori yosetsu no tekiyo

    Energy Technology Data Exchange (ETDEWEB)

    Morita, A [Ishikawajima-Harima Heavy Industries, Co. Ltd., Tokyo (Japan)

    1991-08-01

    Keeping the pace with recent development and application of laser cladding, hard-facing is used more frequently on turbine blades made of superalloys used in aeroengines. This paper explains the basic principles and features of laser hard-facing technique, welding parameters, and examples of practical use. Examples of practical use include application to turbine blades used in ALF502R-5 turbo fan engines for commuter aircraft and high-pressure turbine blades used in RB211 turbo fan engines for large passenger aircraft. In the former engine, improvement of abrasion resistance was intended at the shroud section where blades are in contact with each other, for which inconel was used as the base material and CO-group alloy as the welding material. The welding used a powder supply system with a laser generator oscillating CO{sub 2} at 5 kW and employing a beam collecting mirror plus scanner to attain a beam covering wider width. Faces with higher performance were obtained than by the conventional TIG welding, and the finishing time was decreased largely. 2 refs., 9 figs., 3 tabs.

  15. SHOSPA-MOD, Hot Spot Factors for Fuel and Clad, Hot Channel Factors

    International Nuclear Information System (INIS)

    Amendola, A.

    1982-01-01

    1 - Nature of the physical problem solved: SHOSPA evaluates the hot spot factors for fuel and cladding as well as the hot channel factor as a function of the confidence level. Moreover, it evaluates the probability on n hot subassemblies. The code has been developed with emphasis on sodium cooled fast reactors, but it is applicable to any type of reactors constituted of bundled fuel rods with single phase coolant. An option for plotting is available in this version. 2 - Restrictions on the complexity of the problem: This code is applicable to any type of reactors constituted of fuel rods with single phase coolant

  16. Applications in nuclear data and reactor physics

    International Nuclear Information System (INIS)

    Cullen, D.E.; Muranaka, R.; Schmidt, J.

    1986-01-01

    This book presents the papers given at a conference on reactor kinetics and nuclear data collections. Topics considered at the conference included nuclear data processing, PWR core design calculations, reactor neutron dosimetry, in-core fuel management, reactor safety analysis, transients, two-phase flow, fuel cycles of research reactors, slightly enriched uranium, highly enriched uranium, reactor start-up, computer codes, and the transport of spent fuel elements

  17. Procedures for the medical application of research reactors (Appendix)

    International Nuclear Information System (INIS)

    Nishihara, H.; Kanda, K.

    2004-01-01

    The Kyoto University Reactor (KUR) is one of the four research reactors in Japan that are currently licensed for medical application, in addition to other research purposes. Taking the KUR as an example, legal and other procedures for using research reactors for boron neutron capture therapy (BNCT) are described, which are practiced in accordance with the 'Provisional Guideline Pertaining to Medical Irradiation by Accelerators and/or Reactors, other than defined by the Medical Service Act' of the Science Council of Japan

  18. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  19. Application of the beta particles backscattering technique for determining the thickness of the cladding in nuclear fuels plate

    International Nuclear Information System (INIS)

    Koshimizu, S.; Ferreira, P.I.; Lima, L.F.C.P. de; Vieira, J.M.; Perez, H.E.B.

    1984-01-01

    A prototype of an instalation to measure thickness of cladding and core of nuclear fuels plate using the beta particles backscattering technique is constructed. The method and calibration system is described. The thickness measurements of the cladding and core were done in a natural uranium fuel plate developed at IPEN. The reliability of the method is confirmed by the metalographic measures analysis. (E.G.) [pt

  20. A pellet-clad interaction failure criterion

    International Nuclear Information System (INIS)

    Howl, D.A.; Coucill, D.N.; Marechal, A.J.C.

    1983-01-01

    A Pellet-Clad Interaction (PCI) failure criterion, enabling the number of fuel rod failures in a reactor core to be determined for a variety of normal and fault conditions, is required for safety analysis. The criterion currently being used for the safety analysis of the Pressurized Water Reactor planned for Sizewell in the UK is defined and justified in this paper. The criterion is based upon a threshold clad stress which diminishes with increasing fast neutron dose. This concept is consistent with the mechanism of clad failure being stress corrosion cracking (SCC); providing excess corrodant is always present, the dominant parameter determining the propagation of SCC defects is stress. In applying the criterion, the SLEUTH-SEER 77 fuel performance computer code is used to calculate the peak clad stress, allowing for concentrations due to pellet hourglassing and the effect of radial cracks in the fuel. The method has been validated by analysis of PCI failures in various in-reactor experiments, particularly in the well-characterised power ramp tests in the Steam Generating Heavy Water Reactor (SGHWR) at Winfrith. It is also in accord with out-of-reactor tests with iodine and irradiated Zircaloy clad, such as those carried out at Kjeller in Norway. (author)

  1. 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).

  2. Civilian application of propulsion reactor in Indonesia

    International Nuclear Information System (INIS)

    Djokolelono, M.; Arbie, B.; Lasman, A.N.

    2000-01-01

    It has been learned that to cope with energy requirement in the remote islands and less developed regions of Indonesia, small or very small nuclear reactors producing electricity and/or process heat could be appropriately applied. The barge mounted propulsion power reactors are the attractive examples so far envisioned, and technology information of which is being exposed to the world these last years. The solutions for least maintenance and no on-site refueling, no radioactive discharge, and no radioactive waste to remain in the user country are among the attractions for further deliberations. It has been understood, however, that there are many uncertainties to overcome, especially for the developing countries to introduce this novel application. International acceptance is the most crucial, availability of first-of-the-kind engineering, prototype or reference plant that would prove licensibility in the vendor's country is the second, and economic competitiveness due to very small size is the third among issues to enlighten. The relevant regulations concerning marine nuclear safety, marine transportation, and proliferation of information, as well as international forums to justify the feasibility of related transfer of technology, are the items that the IAEA could help to provide to smoothen any possible international transaction. Indonesia supports this AGM as one of the appropriate IAEA efforts in this line, and expects from it positive international consensus and possible studies/R and D work that this country could participate in. (author)

  3. Application of Bondarenko formalism to fusion reactors

    International Nuclear Information System (INIS)

    Soran, P.D.; Dudziak, D.J.

    1975-01-01

    The Bondarenko formalism used to account for resonance self-shielding effects (temperature and composition) in a Reference Theta-Pinch Reactor is reviewed. A material of interest in the RTPR blanket is 93 Nb, which exhibits a large number of capture resonance in the energy region below 800 keV. Although Nb constitutes a small volume fraction of the blanket, its presence significantly affects the nucleonic properties of the RTPR blanket. The effects of self-shielding in 93 Nb on blanket parameters such as breeding ratio, total afterheat, radioactivity, magnet-coil heating and total energy depositions have been studied. Resonance self-shielding of 93 Nb, as compared to unshielded cross sections, will increase tritium breeding by approximately 7 percent in the RTPR blanket and will decrease blanket radioactivity, total recoverable energy, and magnet-coil heating. Temperature effects change these parameters by less than 2 percent. The method is not restricted to the RTPR, as a single set of Bondarenko f-factors is suitable for application to a variety of fusion reactor designs

  4. 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)

  5. Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

    International Nuclear Information System (INIS)

    Bagley, K.Q.

    1978-01-01

    The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

  6. Applications of plasma core reactors to terrestrial energy systems

    International Nuclear Information System (INIS)

    Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

  7. Application of non-destructive liner thickness measurement technique for manufacturing and inspection process of zirconium lined cladding tube

    International Nuclear Information System (INIS)

    Nakazawa, Norio; Fukuda, Akihiro; Fujii, Noritsugu; Inoue, Koichi

    1986-01-01

    Recently, in order to meet the difference of electric power demand owing to electric power situation, large scale load following operation has become necessary. Therefore, the development of the cladding tubes which withstand power variation has been carried out, as the result, zirconium-lined zircaloy 2 cladding tubes have been developed. In order to reduce the sensitivity to stress corrosion cracking, these zirconium-lined cladding tubes require uniform liner thickness over the whole surface and whole length. Kobe Steel Ltd. developed the nondestructive liner thickness measuring technique based on ultrasonic flaw detection technique and eddy current flaw detection technique. These equipments were applied to the manufacturing and inspection processes of the zirconium-lined cladding tubes, and have demonstrated superiority in the control and assurance of the liner thickness of products. Zirconium-lined cladding tubes, the development of the measuring technique for guaranteeing the uniform liner thickness and the liner thickness control in the manufacturing and inspection processes are described. (Kako, I.)

  8. Application of JAERI research reactors to education

    International Nuclear Information System (INIS)

    Ogawa, Shigeru; Morozumi, Minoru

    1987-01-01

    At the dawning of the atomic age in Japan, training on reactor operation and reactor engineering experiments has been started in 1958 using JRR-1 (a 50 kW water boiler type reactor with liquid fuel), which was the first research reactor in Japan. The role of the training has been transferred to JRR-4 (a 3500 kW swimming pool type reactor with ETR type fuel) since 1969 due to the decommission of JRR-1. The training courses which have been held are: JRR-1 Short-Term Course for Operation (1958 ∼ 1963) General Course (1961 ∼ ) Reactor Engineering Course (1976 ∼ ) Training Course in Nuclear Technology (International course)(1986 ∼ ). And individual training concerning research reactors for the participants of scientist exchange program sponsored by Science and Technology Agency and of bilateral agreement have been initiated in 1985. The graduates of these courses work as staff members in various fields in nuclear industry. (author)

  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. Theoretical studies of the influence of filler material gas gap and cladding material on rewetting rate of nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Blackburn, D.; Pearson, K.G.; Shires, G.L.

    1977-03-01

    Theoretical studies of the effect of fuel and gas gap on the rewetting rate of overheated fuel pins quenched by a falling film of water are presented. Two approaches have been made: a finite difference technique and an approximate analytical solution. The results obtained by the two methods for the case of a uranium-dioxide-filled Zircaloy clad fuel pin are in close agreement. The paper shows that under high pressure conditions the delaying effect of the stored heat within the fuel on the wetting rate is relatively small, particularly if a gas gap is present between the clad and the fuel. At low pressure conditions, however, the effect of the fuel may be very important. Simplification of the analytical solution shows that at low wetting rates a constant fractional reduction in wetting speed may be anticipated the magnitude of which depends only on the relative thermal diffusivities and heat capacities of the fuel and cladding. (author)

  11. Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

    International Nuclear Information System (INIS)

    Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

    1977-01-01

    GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

  12. Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.; Johnson, Gerald D.; Brown, W. F.; Paxton, Michael M.; Puigh, Raymond J.; Eiholzer, Cheryl R.; Martinez, C.; Blotter, M. A.

    2000-02-28

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

  13. Investigation of Y2O3 distribution in the welded joints of the fast reactor fuel claddings made of oxide dispersion strengthened steel

    International Nuclear Information System (INIS)

    Tabakin, E.M.; Kuz'min, S.V.; Ivanovich, Yu.V.; Ukai, Sh.; Kaito, T.; Seki, M.

    2007-01-01

    The study results of Y 2 O 3 distribution in welded joints of claddings from oxide dispersion strengthened steel produced by the technique of powder metallurgy are given in this paper. Change of content and distribution uniformity of yttrium oxide in welds in comparison with metal shell is the result of using flash welding of thin-walled fuel claddings. It is shown that concentration and yttrium oxide distribution uniformity in the cross section of weld, made by pulse laser welding is more high as compared with argon-arc welding [ru

  14. The effect of hot spots upon swelling of Zircaloy cladding as modelled by the code CANSWEL-2

    International Nuclear Information System (INIS)

    Haste, T.J.; Gittus, J.H.

    1980-12-01

    The code CANSWEL-2 models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised-water reactor (PWR). It can treat azimuthal non-uniformities in cladding thickness and temperature, and model the mechanical restraint imposed by the nearest neighbouring rods, including situations where cladding is forced into non-circular shapes. The physical and mechanical models used in the code are presented. Applications of the code are described, both as a stand-alone version and as part of the PWR LOCA code MABEL-2. Comparison with a limited number of relevant out-of-reactor creep strain experiments has generally shown encouraging agreement with the data. (author)

  15. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor; Etude de la corrosion uniforme d'un alliage d'aluminium utilise comme gainage du combustible nucleaire du reacteur experimental Jules Horowitz

    Energy Technology Data Exchange (ETDEWEB)

    Wintergerst, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)

    2008-07-01

    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  16. Application of adjoint sensitivity analysis to nuclear reactor fuel rod performance

    International Nuclear Information System (INIS)

    Wilderman, S.J.; Was, G.S.

    1984-01-01

    Adjoint sensitivity analysis in nuclear fuel behavior modeling is extended to operate on the entire power history for both Zircaloy and stainless steel cladding via the computer codes FCODE-ALPHA/SS and SCODE/SS. The sensitivities of key variables to input parameters are found to be highly non-intuitive and strongly dependent on the fuel-clad gap status and the history of the fuel during the cycle. The sensitivities of five key variables, clad circumferential stress and strain, fission gas release, fuel centerline temperature and fuel-clad gap, to eleven input parameters are studied. The most important input parameters (yielding significances between 1 and 100) are fabricated clad inner and outer radii and fuel radius. The least important significances (less than 0.01) are the time since reactor start-up and fuel-burnup densification rate. Intermediate to these are fabricated fuel porosity, linear heat generation rate, the power history scale factor, clad outer temperature, fill gas pressure and coolant pressure. Stainless steel and Zircaloy have similar sensitivities at start-up but these diverges a burnup proceeds due to the effect of the higher creep rate of Zircaloy which causes the system to be more responsive to changes in input parameters. The value of adjoint sensitivity analysis lies in its capability of uncovering dependencies of fuel variables on input parameters that cannot be determined by a sequential thought process. (orig.)

  17. Application of research reactors for radiation education

    International Nuclear Information System (INIS)

    Ito, Yasuo; Harasawa, Susumu; Hayashi, Shu A.; Tomura, Kenji; Matsuura, Tatsuo; Nakanishi, Tomoko M.; Yamamoto, Yusuke

    1999-01-01

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  18. Application of research reactors for radiation education

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Yasuo [Tokyo Univ. (Japan). Research Center for Nuclear Science and Technology; Harasawa, Susumu; Hayashi, Shu A.; Tomura, Kenji; Matsuura, Tatsuo; Nakanishi, Tomoko M.; Yamamoto, Yusuke

    1999-09-01

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  19. Quantifying reactor safety margins: Application of CSAU [Code Scalability, Applicability and Uncertainty] methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes

    International Nuclear Information System (INIS)

    Wulff, W.; Boyack, B.E.; Duffey, R.B.

    1988-01-01

    Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper. Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PF1/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. 25 refs., 10 figs., 11 tabs

  20. Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1985-01-01

    This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

  1. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    E. Siegmann

    2004-01-01

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  2. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Sattari-Far, Iradi; Andersson, Magnus

    2006-06-01

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the measurement of

  3. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the

  4. Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States)

    2017-11-29

    500 oC, respectively. The coating layer contains both carbon and vanadium elements as quantified by WED, and the phases mainly consist of a mixture of V2C and VC, which was confirmed using X-ray diffraction patterns. In addition, the ratio between V and C varies with processing temperature, and it was observed that a higher temperature promotes the carbon adsorption and increases thickness of the coating. With optimized deposition conditions, we can apply the coating technique toward the actual T91 cladding materials, and provide the possibilities for the real application in sodium-cooled fast reactors (SFRs). Diffusion couple experiments were performed at both 550 oC and 690 oC, which corresponds to normal and aggressive operating temperatures, respectively. The results show that vanadium carbide coating with wider thickness (8 µm) and lower carbon concentration (27 at.%) reduced the width of the inter diffusion region, indicating that vanadium carbide coating can mitigate FCCI effectively. In specific, inter-diffusion between Fe and Ce was prohibited over most area, but Ce diffusion occurred toward the coating and the Fe substrate through thinner coating layer, which needs further optimization in terms of uniform coating thickness. Overall, it is concluded that this coating process can be successfully applied onto the inner surface of HT9 cladding tubes and the FCCI can be effectively mitigated if not totally eliminated.

  5. PRA insights applicable to the design of the Broad Applications Test Reactor

    International Nuclear Information System (INIS)

    Khericha, S.T.; Reilly, H.J.

    1993-01-01

    Design insights applicable to the design of a new Broad Applications Test Reactor (BATR), being studied at Idaho National Engineering Laboratory, are summarized. Sources of design insights include past probabilistic risk assessments and related studies for department of Energy-owned Class A reactors and for commercial reactors. The report includes a preliminary risk allocation scheme for the BATR

  6. Advanced reactor development for non-electric applications

    International Nuclear Information System (INIS)

    Chang, M.H.; Kim, S.H.

    1996-01-01

    Advance in the nuclear reactor technology achieved through nuclear power programs carried out in the world has led nuclear communities to direct its attention to a better and peaceful utilization of nuclear energy in addition to that for power generation. The efforts for non-electric application of nuclear energy has been pursued in a limited number of countries in the world for their special needs. However, those needs and the associated efforts contributed largely to the development and practical realization of advanced reactors characterized by highly improved reactor safety and reliability by deploying the most up-to-date safety technologies. Due mainly to the special purpose of utilization, economic reasons and ease in implementation of new advanced technologies, small and medium reactors have become a major stream in the reactor developments for non-electric applications. The purpose of this paper is to provide, to the interested nuclear society, the overview of the development status and design characteristics of selected advanced nuclear reactors previously developed and/or currently under development specially for non-electric applications. Major design technologies employed in those reactors to enhance the reactor safety and reliability are reviewed to present the underlying principles of the design. Along with the overview, this paper also introduces a development program and major design characteristics of an advanced integral reactor (SMART) for co-generation purpose currently under conceptual development in Korea. (author)

  7. Laser surface cladding of Ti-6Al-4V on AISI 316L stainless steel for bio-implant application

    CSIR Research Space (South Africa)

    Kumar, A

    2014-01-01

    Full Text Available The present study concerns an in-depth investigation of the influence of laser surface cladding of Ti-6Al-4V on the microstructure (both the top surface, cross-section and interface), wear resistance, corrosion resistance and bio-activity of AISI...

  8. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Ruck, G.W.; Lee, A.Y.; Smeltzer, G.; Prevenslik, T.

    1978-11-01

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

  9. Industrial applications of multi-functional, multi-phase reactors

    NARCIS (Netherlands)

    Harmsen, G.J.; Chewter, L.A.

    1999-01-01

    To reveal trends in the design and operation of multi-functional, multi-phase reactors, this paper describes, in historical sequence, three industrial applications of multi-functional, multi-phase reactors developed and operated by Shell Chemicals during the last five decades. For each case, we

  10. Assessment of nuclear reactor concepts for low power space applications

    Science.gov (United States)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  11. Reactor noise analysis applications in NPP I and C systems

    Energy Technology Data Exchange (ETDEWEB)

    Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

    2006-07-01

    Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

  12. Nuclear reactor development in China for non-electrical applications

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Dong Duo; Xu Yuanhui

    1998-01-01

    In parallel to its vigorous program of nuclear power generation, China has attached great importance to the development of nuclear reactors for non-electrical applications. The Institute of Nuclear Energy Technology (INET) in Beijing has been developing technologies of the water-cooled heating reactor and the modular high temperature gas-cooled reactor. In 1989, a 5 MW water cooled test reactor was erected. Currently, an industrial demonstration nuclear heating plant is being projected. Feasibility studies are being made of sea-water desalination using the INET developed nuclear heating reactor as heat source. Also, a 10 MW high temperature gas-cooled test reactor is being constructed at INET in the framework of China's national high-tech program. The paper gives an overview of China's energy market situation. With respect to China's technology development of high temperature gas-cooled reactors and water cooled heating reactors, the paper describes some general requirements on the technical development, reviews the national programs and activities, describes briefly the design and safety features of the reactor concepts, discusses aspects of application potentials. (author)

  13. 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)

  14. PRA insights applicable to the design of a broad applications test reactor

    International Nuclear Information System (INIS)

    Khericha, S.T.; Reilly, H.J.

    1993-01-01

    Design insights applicable to the design of a new Broad Applications Test Reactor (BATR), studied during Fiscal Years 1992 an d1993 at Idaho National Engineering Laboratory (INEL), are summarized. Sources of design insights include past probabilistic risk assessments (PRAs) and related studies for Department of Energy (DOE)-owned Class A reactors and for commercial reactors. The report includes preliminary risk allocations for the BATR. The survey addressed those design insights that would affect the reactor core damage frequency (CDF). The design insights, while selected specifically for BATR, should be applicable to any new advanced test reactor

  15. 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

  16. Artificial intelligence program in a computer application supporting reactor operations

    International Nuclear Information System (INIS)

    Stratton, R.C.; Town, G.G.

    1985-01-01

    Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II

  17. Application of autoregressive moving average model in reactor noise analysis

    International Nuclear Information System (INIS)

    Tran Dinh Tri

    1993-01-01

    The application of an autoregressive (AR) model to estimating noise measurements has achieved many successes in reactor noise analysis in the last ten years. The physical processes that take place in the nuclear reactor, however, are described by an autoregressive moving average (ARMA) model rather than by an AR model. Consequently more correct results could be obtained by applying the ARMA model instead of the AR model to reactor noise analysis. In this paper the system of the generalised Yule-Walker equations is derived from the equation of an ARMA model, then a method for its solution is given. Numerical results show the applications of the method proposed. (author)

  18. Influence of specimen design on the ductility of zircaloy cladding: Experiment and analysis

    International Nuclear Information System (INIS)

    Bates, D. W.; Majumdar, S.; Koss, D. A.; Motta, A. T.

    1999-01-01

    In a reactivity-initiated accident (RIA), a control rod ejection or drop causes a sudden increase in reactor power, which in turn deposits a large amount of energy into the fuel. The resulting thermal expansion and fission gas release loads the cladding into the plastic regime and may cause it to fail. In order to predict cladding survivability, there has been considerable interest and effort in supplementing integral WA tests with separate-effects ring tests of cladding tubes. Such tests can give one insight into failure mechanisms and measure relevant mechanical properties (such as yield strength, uniform elongation, uniaxial stress-strain curve, etc.), for use in computer codes that attempt to predict cladding response during an RIA. The accuracy of such model predictions obviously depends on appropriate and accurate failure data. This study concerns itself with the proper development of ring tensile tests that (i) are similar to the loading conditions present in an RIA, (ii) measure the relevant mechanical properties and (iii) provide insight regarding the influence of the strain paths on the failure mechanisms present if Zircaloy cladding. Based on both experiments and computational modeling, the authors investigate the failure of Zircaloy tubing as a function of specimen geometry, and discuss the limitations of certain ring-test geometries in yielding failure ductility data that are applicable to RIA situations

  19. Fracture assessment of weld material from a full-thickness clad RPV shell segment

    International Nuclear Information System (INIS)

    Keeney, J.A.; Bass, B.R.; McAfee, W.J.

    1996-01-01

    Fracture analysis was applied to full-thickness clad beam specimens containing shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPV) at beginning of life. The beam specimens were fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include gradients of material properties and residual stresses due to welding and cladding applications. Fracture toughness estimates were obtained from load vs load-line displacement and load vs crack-mouth-opening displacement data using finite-element methods and estimation schemes based on the η-factor method. One of the beams experienced a significant amount of precleavage stable ductile tearing. Effects of precleavage tearing on estimates of fracture toughness were investigated using continuum damage models. Fracture toughness results from the clad beam specimens were compared with other deep- and shallow-crack single-edge notch bend (SENB) data generated previously from A533 Grade B plate material. Range of scatter for the clad beam data is consistent with that from the laboratory-scale SENB specimens tested at the same temperature

  20. Study of reactions between fuel (mixed oxide (UPu)Osub(2-x)) and cladding (stainless-steel) in reactors: influence of iodine compounds

    International Nuclear Information System (INIS)

    Aubert, Michel.

    1976-03-01

    The influence of iodine compounds on the development of the oxide-cladding reaction was examined. The action of iodine, cesium and cesium iodide on type 316 stainless was determined in the presence or absence of uranium oxide or mixed uranium-plutonium oxide type fuel in a closed system, isothermal or with a temperature gradient. The study of the stainless steel iodine reactions was developed in particular. These experiments showed that cesium combines with uranium oxide to give cesium uranate Cs 2 U 2 O 7 ; it is not unreasonable to suppose that cesium urano-plutonate Cs 2 (U,Pu) 2 O 7 could be formed inside the pile. It was then shown that cesium iodide in the presence of sufficiently non-stoichiometric mixed oxide could contribute towards the degradation of the stainless steel cladding. Under these conditions the reaction is accompained by a transport of manganese, chromium and iron into the hot parts of the fuel by a Van-Arkel type mechanism. This might explain the presence of metallic precipitates in the fuel, but the role assigned to molybdenum iodide in the same phenomenon is considered unlikely. Finally it is proposed to deposit a thin layer of manganese metal on the inner surface of the cladding in order to minimize the action of fission products (CsI, Te) [fr

  1. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1976-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

  2. RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

    International Nuclear Information System (INIS)

    Hohorst, J.K.; Allison, C.M.

    2010-01-01

    The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

  3. RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)

    2010-07-01

    The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

  4. Prevention of nuclear fuel cladding materials corrosion

    International Nuclear Information System (INIS)

    Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

    1983-01-01

    The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

  5. Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yurko, Joseph P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swindlehurst, Gregg [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

  6. Study on the application of CANDLE burnup strategy to several nuclear reactors. JAERI's nuclear research promotion program, H13-002 (Contract research)

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko

    2005-03-01

    The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. When this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means that the nuclear energy can be utilized for many hundreds years without new mining, enrichment and reprocessing, and the amount of spent fuel can be reduced considerably. However, in order to perform such a high fuel burnup some innovative technologies should be developed. Though development of innovative fuel will take a lot of time, intermediate re-cladding may be easy to be employed. Compared to fast reactors, application of CANDLE burnup to prismatic fuel high-temperature gas cooled reactors is very easy. In this report the application of CANDLE burnup to both these types of reactors are studied. (author)

  7. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  8. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    Arzhanov, V.

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  9. Incidental transients problems in reactor. Application examples

    International Nuclear Information System (INIS)

    Marbach, G.

    1988-03-01

    The fast neutron reactor fuel element qualification should be made not only for nominal operation but also for incidental and accidental transients. Different studies and tests permit to bring this justification such as simulation in hot laboratory after irradiation of irradiated pins or specific tests interpretation [fr

  10. Processing of nuclear data for reactor applications

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.

    1996-01-01

    A brief description is given of the processing and validation of nuclear data in connection with the TRX-1, TRX-2, BAPL-1 and BAPL-2 benchmarks of a/o thermal reactors and in connection with the JEF-1, JENDL-3 and WIMS libraries. Also, the validation of the WLUP results are briefly discussed. 8 refs, 5 tabs

  11. Engineered zircaloy cladding modifications for improved accident tolerance of LWR fuel: US DOE NEUP Integrated Research Project

    International Nuclear Information System (INIS)

    Heuser, Brent

    2013-01-01

    An integrated research project (IRP) to fabricate and evaluate modified zircaloy LWR cladding under normal BWR/PWR operation and off-normal events has been funded by the US DOE. The IRP involves three US academic institutions, a US national laboratory, an intermediate stock industrial cladding supplier, and an international academic institution. A combination of computational and experimental protocols will be employed to design and test modified zircaloy cladding with respect to corrosion and accelerated oxide growth, the former associated with normal operation, the latter associated with steam exposure during loss of coolant accidents (LOCAs) and low-pressure core re-floods. Efforts will be made to go beyond design-base accident (DBA) scenarios (cladding temperature equal to or less than 1204 deg. C) during the experimental phase of modified zircaloy performance characterisation. The project anticipates the use of the facilities at ORNL to achieve steam exposure beyond DBA scenarios. In addition, irradiation of down-selected modified cladding candidates in the ATR may be performed. Cladding performance evaluation will be incorporated into a reactor system modelling effort of fuel performance, neutronics, and thermal hydraulics, thereby providing a holistic approach to accident-tolerant nuclear fuel. The proposed IRP brings together personnel, facilities, and capabilities across a wide range of technical areas relevant to the study of modified nuclear fuel and LWR performance during normal operation and off-normal scenarios. Two pathways towards accident-tolerant LWR fuel are envisioned, both based on the modification of existing zircaloy cladding. The first is the modification of the cladding surface by the application of a coating layer designed to shift the M + O→MO reaction away from oxide growth during steam exposure at elevated temperatures. This pathway is referred to as the 'surface coating' solution. The second is the modification of the bulk

  12. Creep behavior of materials for high-temperature reactor application

    International Nuclear Information System (INIS)

    Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

    1984-01-01

    Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

  13. Chemical compatibility between UO{sub 2} fuel and SiC cladding for LWRs. Application to ATF (Accident-Tolerant Fuels)

    Energy Technology Data Exchange (ETDEWEB)

    Braun, James, E-mail: james.braun@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Guéneau, Christine; Alpettaz, Thierry [DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Sauder, Cédric [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Brackx, Emmanuelle; Domenger, Renaud [CEA, DEN, Marcoule, Metallography and Chemical Analysis Laboratory, F-30207 Bagnols-sur-Cèze (France); Gossé, Stéphane [DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Balbaud-Célérier, Fanny [DEN-Service d’Etudes Analytiques et de Réactivité des Surfaces (SEARS), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2017-04-15

    Silicon carbide-silicon carbide (SiC/SiC) composites are considered to replace the current zirconium-based cladding materials thanks to their good behavior under irradiation and their resistance under oxidative environments at high temperature. In the present work, a thermodynamic analysis of the UO{sub 2±x}/SiC system is performed. Moreover, using two different experimental methods, the chemical compatibility of SiC towards uranium dioxide, with various oxygen contents (UO{sub 2±x}) is investigated in the 1500–1970 K temperature range. The reaction leads to the formation of mainly uranium silicides and carbides phases along with CO and SiO gas release. Knudsen Cell Mass Spectrometry is used to measure the gas release occurring during the reaction between UO{sub 2+x} and SiC powders as function of time and temperature. These experimental conditions are representative of an open system. Diffusion couple experiments with pellets are also performed to study the reaction kinetics in closed system conditions. In both cases, a limited chemical reaction is observed below 1700 K, whereas the reaction is enhanced at higher temperature due to the decomposition of SiC leading to Si vaporization. The temperature of formation of the liquid phase is found to lie between 1850 < T < 1950 K. - Highlights: •A limited chemical reaction occurs between SiC and UO{sub 2+x} up to 1514 K. •CO gas along with the generation of USi{sub x} are detected over 1514 K in open system. •A liquid phase forms between 1850 and 1950 K in the UO{sub 2+x}/SiC system. •Results are encouraging for the use of SiC/SiC cladding in nuclear reactors.

  14. Effects of alloying and processing modifications on precipitation and strength in 9%Cr ferritic/martensitic steels for fast reactor cladding

    Science.gov (United States)

    Tippey, Kristin E.

    P92 was modified with respect to alloying and processing in the attempt to enhance high-temperature microstructural stability and mechanical properties. Alloying effects were modeled in ThermoCalcRTM and analyzed with reference to literature. ThermoCalcRTM modeling was conducted to design two low-carbon P92-like low-carbon alloys with austenite stabilized by alternative alloying; full conversion to austenite allows for a fully martensitic structure. Goals included avoidance of Z-phase, decrease of M23C6 phase fraction and maintained or increased MX phase fraction. Fine carbonitride precipitation was optimized by selecting alloying compositions such that all V and Nb could be solutionized at temperatures outside the delta-ferrite phase field. A low-carbon alloy (LC) and a low-carbon-zero-niobium alloy (0Nb) were identified and fabricated. This low-carbon approach stems from the increased creep resistance reported in several low-carbon alloys, presumably from reduced M23C6 precipitation and maintained MX precipitation [1], although these low-carbon alloys also contained additional tungsten (W) and cobalt (Co) compared to the base P92 alloy. The synergistic effect of Co and W on the microstructure and mechanical properties are difficult to deconvolute. Higher solutionizing temperatures allow more V and Nb into solution and increase prior austenite grain size; however, at sufficiently high temperatures delta-ferrite forms. Optimal solutionizing temperatures to maximize V and Nb in solution, while avoiding the onset of the delta ferrite phase field, were analyzed in ThermoCalcRTM. Optical microscopy showed ThermoCalc RTM predicted higher delta-ferrite onset temperatures of 20 °C in P92 alloys to nearly 50 °C in the designed alloys of the critical temperature. Identifying the balance where maximum fine precipitation is achieved and delta-ferrite avoided is a key factor in the design of an acceptable P92-like alloy for Generation IV reactor cladding. Processing was

  15. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  16. Reactor coolant pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  17. Management of cladding hulls and fuel hardware

    International Nuclear Information System (INIS)

    1985-01-01

    The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

  18. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-01-01

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

  19. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

    2013-10-15

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  20. Superalloy applications in the fast breeder reactor

    International Nuclear Information System (INIS)

    Powell, R.W.

    1976-01-01

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

  1. Application of SAFE to an operating reactor

    International Nuclear Information System (INIS)

    Chapman, L.D.

    1979-01-01

    A method for the evaluation of physical protection systems at nuclear facilities has been developed. The evaluation process consists of five major phases: (1) Facility Characterization, (2) Facility Representation, (3) Component Performance, (4) Adversary Path Analysis, and (5) Effectiveness Evaluation. Each of these phases will be described in some detail and illustrated by examples. The process for evaluation of physical protection system effectiveness against an outside threat will be presented for a reactor facility

  2. 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....

  3. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

  4. Preliminary assessment of the fracture behavior of weld material in full-thickness clad beams

    International Nuclear Information System (INIS)

    Keeney, J.A.; Bass, B.R.; McAfee, W.J.; Iskander, S.K.

    1994-10-01

    This report describes a testing program that utilizes full-thickness clad beam specimens to quantify fracture toughness for shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include material gradients due to welding and cladding applications, as well as material inhomogeneities in welded regions due to reheating in multiple weld passes. A summary of the testing program includes a description of the specimen geometry, material properties, the testing procedure, and the experimental results form three specimens. The yield strength of the weld material was determined to be 36% higher than the yield strength of the base material. An irradiation-induced increase in yield strength of the weld material could result in a yield stress that exceeds the upper limit where code curves are valid. The high yield strength for prototypic weld material may have implications for RPV structural integrity assessments. Analyses of the test data are discussed, including comparisons of measured displacements with finite-element analysis results, applications of toughness estimation techniques, and interpretations of constraint conditions implied by stress-based constraint methodologies. Metallurgical conditions in the region of the cladding heat-affected zone are proposed as a possible explanation for the lower-bound fracture toughness measured with one of the shallow-crack clad beam specimens. Fracture toughness data from the three clad beam specimens are compared with other shallow- and deep-crack uniaxial beam and cruciform data generated previously from A 533 Grade B plate material

  5. Laser cladding of quasicrystalline alloys

    International Nuclear Information System (INIS)

    Audebert, F.; Sirkin, H.; Colaco, R.; Vilar, R.

    1998-01-01

    Quasicrystals are a new class of ordinated structures with metastable characteristics room temperature. Quasicrystalline phases can be obtained by rapid quenching from the melt of some alloys. In general, quasicrystals present properties which make these alloys promising for wear and corrosion resistant coatings applications. During the last years, the development of quasicrystalline coatings by means of thermal spray techniques has been impulsed. However, no references have been found of their application by means of laser techniques. In this work four claddings of quasicrystalline compositions formed over aluminium substrate, produced by a continuous CO 2 laser using simultaneous powders mixture injection are presented. The claddings were characterized by X ray diffraction, scanning electron microscopy and Vickers microhardness. (Author) 18 refs

  6. Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2011-08-15

    Highlights: > Microstructural changes in stainless steel electroslag weld-overlay cladding. > Thermal aging caused progress of spinodal decomposition and precipitation of G phases in the {delta}-ferrite phase. > The degree of the spinodal decomposition had a linear relationship to the hardness. - Abstract: The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% {delta}-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 deg. C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M{sub 23}C{sub 6} type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.

  7. 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

  8. 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.)

  9. The Virtual Environment for Reactor Applications (VERA): Design and architecture

    International Nuclear Information System (INIS)

    Turner, John A.; Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin; Pawlowski, Roger; Schmidt, Rodney; Summers, Randall

    2016-01-01

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.

  10. The Virtual Environment for Reactor Applications (VERA): Design and architecture

    Energy Technology Data Exchange (ETDEWEB)

    Turner, John A., E-mail: turnerja@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Pawlowski, Roger; Schmidt, Rodney; Summers, Randall [Sandia National Laboratories, Albuquerque, NM 87185 (United States)

    2016-12-01

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.

  11. Ceramics as nuclear reactor fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    1975-01-01

    Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

  12. Organic materials for fusion-reactor applications

    International Nuclear Information System (INIS)

    Hurley, G.F.; Coltman, R.R. Jr.

    1983-09-01

    Organic materials requirements for fusion-reactor magnets are described with reference to the temperature, radiation, and electrical and mechanical stress environment expected in these magnets. A review is presented of the response to gamma-ray and neutron irradiation at low temperatures of candidate organic materials; i.e. laminates, thin films, and potting compounds. Lifetime-limiting features of this response as well as needed testing under magnet operating conditions not yet adequately investigated are identified and recomendations for future work are made

  13. Westinghouse small modular reactor design and application

    Energy Technology Data Exchange (ETDEWEB)

    Blinn, R.; Godfrey, M. [Westinghouse Electric Company, Cranberry Township, Pennsilvania (United States)

    2012-07-01

    The AP1000 is currently under construction in both China and the US with the first one scheduled to come on line in late 2013. Nuclear power is a proven, safe, plentiful and clean source of power generation, and Westinghouse Electric Company, the pioneer and global leader in nuclear plant design and construction, is ready with the AP1000™ pressurized water reactor (PWR). The AP1000, based on the proven performance of Westinghouse-designed PWRs, is an advanced 1154 MWe nuclear power plant that uses the forces of nature and simplicity of design to enhance plant safety and operations and reduce construction costs.

  14. Ceramic Coatings for Clad (The C3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sickafus, Kurt E. [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Miller, Larry [Univ. of Tennessee, Knoxville, TN (United States); Weber, Bill [Univ. of Tennessee, Knoxville, TN (United States); Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States); Patel, Maulik [Univ. of Tennessee, Knoxville, TN (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Wolfe, Doug [Pennsylvania State Univ., University Park, PA (United States); Fratoni, Max [Univ. of California, Berkeley, CA (United States); Raj, Rishi [Univ. of Colorado, Boulder, CO (United States); Plunkett, Kenneth [Univ. of Colorado, Boulder, CO (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Hollis, Kendall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Chris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Comstock, Robert [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Partezana, Jonna [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Whittle, Karl [Univ. of Sheffield (United Kingdom); Preuss, Michael [Univ. of Manchester (United Kingdom); Withers, Philip [Univ. of Manchester (United Kingdom); Wilkinson, Angus [Univ. of Oxford (United Kingdom); Donnelly, Stephen [Univ. of Huddersfield (United Kingdom); Riley, Daniel [Australian Nuclear Science and Technology Organisation, Syndney (Australia)

    2017-02-14

    The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectives of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as

  15. Fuel cladding mechanical properties for transient analysis

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

    1976-01-01

    Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

  16. Potential effects of gallium on cladding materials

    International Nuclear Information System (INIS)

    Wilson, D.F.; Beahm, E.C.; Besmann, T.M.; DeVan, J.H.; DiStefano, J.R.; Gat, U.; Greene, S.R.; Rittenhouse, P.L.; Worley, B.A.

    1997-10-01

    This paper identifies and examines issues concerning the incorporation of gallium in weapons derived plutonium in light water reactor (LWR) MOX fuels. Particular attention is given to the more likely effects of the gallium on the behavior of the cladding material. The chemistry of weapons grade (WG) MOX, including possible consequences of gallium within plutonium agglomerates, was assessed. Based on the calculated oxidation potentials of MOX fuel, the effect that gallium may have on reactions involving fission products and possible impact on cladding performance were postulated. Gallium transport mechanisms are discussed. With an understanding of oxidation potentials and assumptions of mechanisms for gallium transport, possible effects of gallium on corrosion of cladding were evaluated. Potential and unresolved issues and suggested research and development (R and D) required to provide missing information are presented

  17. 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

  18. Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

    Science.gov (United States)

    Narukawa, Takafumi; Yamaguchi, Akira; Jang, Sunghyon; Amaya, Masaki

    2018-02-01

    For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions of light-water-reactors, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best among the three models to estimate the fracture probability in terms of the degree of prediction accuracy for both next data to be obtained and the true model. Using the log-probit model, it was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.

  19. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    Doca, Cezar

    2001-01-01

    By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

  20. 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.

  1. IAEA activities supporting the applications of research reactors in 2013

    International Nuclear Information System (INIS)

    Peld, N.D.; Ridikas, D.

    2014-01-01

    As the underutilization of research reactors around the world persists as a primary topic of concern among facility owners and operators, the IAEA responded in 2013 with a broad range of activities to address the planning, execution and improvement of many experimental techniques. The revision of two critical documents for planning and diversifying a facility's portfolio of applications, TECDOC 1234 'The Applications of Research Reactors' and TECDOC 1212 'Strategic Planning for Research Reactors', is in progress in order to keep this information relevant, corresponding to the dynamism of experimental techniques and research capabilities. Related to the latter TECDOC, the IAEA convened a meeting in 2013 for the expert review of a number of strategic plans submitted by research reactor operators in developing countries. A number of activities focusing on specific applications are either continuing or beginning as well. In neutron activation analysis, a joint round of inter-comparison proficiency testing sponsored by the IAEA Technical Cooperation Department will be completed, and facility progress in measurement accuracy is described. Also, a training workshop in neutron imaging and Coordinated Research Projects in reactor benchmarks, automation of neutron activation analysis and neutron beam techniques for material testing intend to advance these activities as more beneficial services to researchers and other users. (author)

  2. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  3. Features of an emergency heat-conducting path in reactors about lead-bismuth and lead heat-carriers

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Bokova, T.A.; Molodtsov, A.A.

    2006-01-01

    The reactor emergency heat removal systems should transfer heat from the surface of reactor core fuel element claddings to the primary circuit followed by heat transfer to the environment. One suggests three design approaches for emergency heat removal systems in lead-bismuth and lead cooled reactor circuits that take account of the peculiar nature of their features. Application of the discussed systems for emergency heat removal improves safety of lead-bismuth and lead cooled reactor plants [ru

  4. Preliminary design studies on the Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Terry, W.J.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

    1992-08-01

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

  5. Nuclear reactor application for high temperature power industrial processes

    International Nuclear Information System (INIS)

    Dollezhal', N.A.; Zaicho, N.D.; Alexeev, A.M.; Baturov, B.B.; Karyakin, Yu.I.; Nazarov, E.K.; Ponomarev-Stepnoj, N.N.; Protzenko, A.M.; Chernyaev, V.A.

    1977-01-01

    This report gives the results of considerations on industrial heat and technology processes (in chemistry, steelmaking, etc.) from the point of view of possible ways, technical conditions and nuclear safety requirements for the use of high temperature reactors in these processes. Possible variants of energy-technological diagrams of nuclear-steelmaking, methane steam-reforming reaction and other processes, taking into account the specific character of nuclear fuel are also given. Technical possibilities and economic conditions of the usage of different types of high temperature reactors (gas cooled reactors and reactors which have other means of transport of nuclear heat) in heat processes are examined. The report has an analysis of the problem, that arises with the application of nuclear reactors in energy-technological plants and an evaluation of solutions of this problem. There is a reason to suppose that we will benefit from the use of high temperature reactors in comparison with the production based on high quality fossil fuel [ru

  6. Improved Nuclear Reactor and Shield Mass Model for Space Applications

    Science.gov (United States)

    Robb, Kevin

    2004-01-01

    New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

  7. ASTEC applications to VVER-440/V213 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: ivstt@nextra.sk; Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir

    2014-06-01

    Since the beginning of ASTEC development by IRSN and GRS the code was widely applied to VVER reactors. In this paper, at first specific features of VVER-440/V213 reactor design that are important from the modelling point of view are briefly described. Then the validation of ASTEC code with focus on its applicability to VVER reactors is briefly summarised and the results obtained with the ASTEC V2.0-rev1 version for the ISP-33 PACTEL natural circulation experiment are presented. In the next section the application of ASTEC V2.0-rev1 code in upgrade of VVER-440/V213 NPPs to cope with consequences of severe accidents is described. This upgrade includes adoption of in-vessel retention via external reactor vessel cooling and installation of large capacity passive autocatalytic recombiners. Results of analysis with focus on corium localisation and stabilisation inside reactor vessel, hydrogen control in confinement and prevention of long-term confinement pressurisation are presented.

  8. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    Directory of Open Access Journals (Sweden)

    Bo Cheng

    2016-02-01

    discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

  9. Achilles tests finally nail PWR fuel clad ballooning fears

    International Nuclear Information System (INIS)

    Dore, P.; McMinn, K.

    1992-01-01

    A conclusive series of experiments carried out by AEA Reactor Services at its Achilles rig in the UK has finally allayed fears that fuel clad ballooning is a major safety problem for Sizewell B, Britain's first Pressurized Water Reactor. The experiments are described in this article. (author)

  10. Cladding axial elongation models for FRAP-T6

    International Nuclear Information System (INIS)

    Shah, V.N.; Carlson, E.R.; Berna, G.A.

    1983-01-01

    This paper presents a description of the cladding axial elongation models developed at the Idaho National Engineering Laboratory (INEL) for use by the FRAP-T6 computer code in analyzing the response of fuel rods during reactor transients in light water reactors (LWR). The FRAP-T6 code contains models (FRACAS-II subcode) that analyze the structural response of a fuel rod including pellet-cladding-mechanical-interaction (PCMI). Recently, four models were incorporated into FRACAS-II to calculate cladding axial deformation: (a) axial PCMI, (b) trapped fuel stack, (c) fuel relocation, and (d) effective fuel thermal expansion. Comparisons of cladding axial elongation measurements from two experiments with the corresponding FRAP-T6 calculations are presented

  11. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

    2017-04-15

    Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

  12. 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...

  13. Vanadium-base alloys for fusion reactor applications

    International Nuclear Information System (INIS)

    Smith, D.L.; Loomis, B.A.; Diercks, D.R.

    1984-10-01

    Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined

  14. Vanadium-base alloys for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Loomis, B.A.; Diercks, D.R.

    1984-10-01

    Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined.

  15. An overview of the HSST Full-Thickness Shallow-Crack Clad Beam Testing Program

    International Nuclear Information System (INIS)

    Keeney, J.A.; Theiss, T.J.; McAfee, W.J.; Bass, B.R.

    1994-01-01

    A testing program is described that will utilize full-thickness clad beam specimens to quantify fracture toughness for shallow flaws in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from a section of an RPV wall that includes weld, plate and clad material. Metallurgical factors potentially influencing fracture toughness for shallow flaws in the beam specimen include material gradients due to welding and cladding applications, as well as material inhomogeneities in welded regions due to reheating in multiple weld passes. Fracture toughness tests focusing on shallow flaws in plate and weld material will also provide data for evaluating the relative influence of absolute and normalized crack depth on constraint conditions. Pretest finite-element analyses are described that provide near-tip stress and strain fields for characterization of constraint in the shallow-crack specimens in terms of the Q-stress. Analysis results predict a constraint loss in the shallow-crack clad beam specimen similar to that determined for a previously tested shallow-crack single-edge notch homogeneous bend specimen with the same normalized crack depth

  16. Improvement technique of sensitized HAZ by GTAW cladding applied to a BWR power plant

    International Nuclear Information System (INIS)

    Tujimura, Hiroshi; Tamai, Yasumasa; Furukawa, Hideyasu; Kurosawa, Kouichi; Chiba, Isao; Nomura, Keiichi.

    1995-01-01

    A SCC(Stress Corrosion Cracking)-resistant technique, in which the sleeve installed by expansion is melted by GTAW process without filler metal with outside water cooling, was developed. The technique was applied to ICM (In-Core Monitor) housings of a BWR power plant in 1993. The ICM housings of which materials are type 304 Stainless Steels are sensitized with high tensile residual stresses by welding to the RPV (Reactor Pressure Vessel). As the result, ICM housings have potential of SCC initiation. Therefore, the improvement technique resistant to SCC was needed. The technique can improve chemical composition of the housing inside and residual stresses of the housing outside at the same time. Sensitization of the housing inner surface area is eliminated by replacing low-carbon with proper-ferrite microstructure clad. High tensile residual stresses of housing outside surface area is improved into compressive side. Compressive stresses of outside surface are induced by thermal stresses which are caused by inside cladding with outside water cooling. The clad is required to be low-carbon metal with proper ferrite and not to have the new sensitized HAZ (Heat Affected Zone) on the surface by cladding. The effect of the technique was qualified by SCC test, chemical composition check, ferrite content measurement and residual stresses measurement etc. All equipment for remote application were developed and qualified, too. The technique was successfully applied to a BWR plant after sufficient training

  17. 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

  18. Analysis of pellet cladding mechanical interaction using computational simulation

    Energy Technology Data Exchange (ETDEWEB)

    Berretta, José R.; Suman, Ricardo B.; Faria, Danilo P.; Rodi, Paulo A., E-mail: jose.berretta@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), São Paulo, SP (Brazil). Laboratório de Análise, Avaliação e Gerenciamento de Riscos

    2017-07-01

    During the operation of Pressurized Water Reactors (PWR), specifically under power transients, the fuel pellet experiences many phenomena, such as swelling and thermal expansion. These dimensional changes in the fuel pellet can enable occurrence of contact it and the cladding along the fuel rod. Thus, pellet cladding mechanical interaction (PCMI), due this contact, induces stress increase at the contact points during a period, until the accommodation of the cladding to the stress increases. This accommodation occurs by means of the cladding strain, which can produce failure, if the fuel rod deformation is permanent or the burst limit of the cladding is reached. Therefore, the mechanical behavior of the cladding during the occurrence of PCMI under power transients shall be investigated during the fuel rod design. Considering the Accident Tolerant Fuel program which aims to develop new materials to be used as cladding in PWR, one important design condition to be evaluated is the cladding behavior under PCMI. The purpose of this paper is to analyze the effects of the PCMI on a typical PWR fuel rod geometry with stainless steel cladding under normal power transients using computational simulation (ANSYS code). The PCMI was analyzed considering four geometric situations at the region of interaction between pellet and cladding. The first case, called “perfect fuel model” was used as reference for comparison. In the second case, it was considered the occurrence of a pellet crack with the loss of a chip. The goal for the next two cases was that a pellet chip was positioned into the gap of pellet-cladding, in the situations described in the first two cases. (author)

  19. Modular helium reactor for non-electric applications

    International Nuclear Information System (INIS)

    Shenoy, A.

    1997-01-01

    The high temperature gas-cooled Modular Helium Reactor (MHR) is an advanced, high efficiency reactor system which can play a vital role in meeting the future energy needs of the world by contributing not only to the generation of electric power, but also the non-electric energy traditionally served by fossil fuels. This paper summarizes work done over 20 years, by several people at General Atomics, how the Modular Helium Reactor can be integrated to provide different non-electric applications during Process Steam/Cogeneration for industrial application, Process Heat for transportation fuel development and Hydrogen Production for various energy applications. The MHR integrates favorably into present petrochemical and primary metal process industries, heavy oil recovery, and future shale oil recovery and synfuel processes. The technical fit of the Process Steam/Cogeneration Modular Helium Reactor (PS/C-MHR) into these processes is excellent, since it can supply the required quantity and high quality of steam without fossil superheating. 12 refs, 25 figs, 2 tabs

  20. WWER water chemistry related to fuel cladding behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez (Czech Republic); Vrtilkova, V [Nuclear Fuel Inst., Prague (Czech Republic)

    1997-02-01

    Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. (author). 14 refs, 5 figs, 3 tabs.

  1. Second phase precipitation in irradiated Type 316 stainless steel cladding

    International Nuclear Information System (INIS)

    Hales, J.W.

    1978-05-01

    Differences in the phase composition of FFTF fuel cans following irradiation in the General Electric Test Reactor compared to HEDL fuel cans prompted laboratory studies to be conducted using cladding from the same lots of material used to fabricate the fuel pins and on cladding sections removed from the plenum area of the irradiated fuel pins to help establish the cause of the observed differences

  2. Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

    Science.gov (United States)

    Takeuchi, T.; Kameda, J.; Nagai, Y.; Toyama, T.; Nishiyama, Y.; Onizawa, K.

    2011-08-01

    The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% δ-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M 23C 6 type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.

  3. Rectangular-cladding silicon slot waveguide with improved nonlinear performance

    Science.gov (United States)

    Huang, Zengzhi; Huang, Qingzhong; Wang, Yi; Xia, Jinsong

    2018-04-01

    Silicon slot waveguides have great potential in hybrid silicon integration to realize nonlinear optical applications. We propose a rectangular-cladding hybrid silicon slot waveguide. Simulation result shows that, with a rectangular-cladding, the slot waveguide can be formed by narrower silicon strips, so the two-photon absorption (TPA) loss in silicon is decreased. When the cladding material is a nonlinear polymer, the calculated TPA figure of merit (FOMTPA) is 4.4, close to the value of bulk nonlinear polymer of 5.0. This value confirms the good nonlinear performance of rectangular-cladding silicon slot waveguides.

  4. International Conference on Physics and Technology of Reactors and Applications

    International Nuclear Information System (INIS)

    2007-01-01

    The first international conference on physics and technology of reactors and applications (PHYTRA 1) which took place in Marrakech (Morocco) from 14 to 16 March 2007, was designed to bring together scientists, teachers and students from universities, research centres and industry and other institutions to exchange knowledge and to discuss ideas and future issues. The programmes of the PHYTRA 1 conference covers a wide variety topics, the conference was organised in three plenary sessions, ten oral technical sessions and two poster sessions. The plenary sessions covers the following topics : The prospects of nuclear energy, The situation of nuclear sciences and energy in Morocco and Africa, and the new development in reactor physics and reactor design [fr

  5. Cermet-fueled reactors for multimegawatt space power applications

    International Nuclear Information System (INIS)

    Cowan, C.L.; Armijo, J.S.; Kruger, G.B.; Palmer, R.S.; Van Hoomisson, J.E.

    1988-01-01

    The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

  6. A gamma heating calculation methodology for research reactor application

    International Nuclear Information System (INIS)

    Lee, Y.K.; David, J.C.; Carcreff, H.

    2001-01-01

    Gamma heating is an important issue in research reactor operation and fuel safety. Heat deposition in irradiation targets and temperature distribution in irradiation facility should be determined so as to obtain the optimal irradiation conditions. This paper presents a recently developed gamma heating calculation methodology and its application on the research reactors. Based on the TRIPOLI-4 Monte Carlo code under the continuous-energy option, this new calculation methodology was validated against calorimetric measurements realized within a large ex-core irradiation facility of the 70 MWth OSIRIS materials testing reactor (MTR). The contributions from prompt fission neutrons, prompt fission γ-rays, capture γ-rays and inelastic γ-rays to heat deposition were evaluated by a coupled (n, γ) transport calculation. The fission product decay γ-rays were also considered but the activation γ-rays were neglected in this study. (author)

  7. JENDL-4.0 benchmarking for fission reactor applications

    International Nuclear Information System (INIS)

    Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

    2011-01-01

    Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

  8. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur de tubes de gainage du combustible des REP en alliage de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, C. [CEA Saclay, Dept. de Recherche sur l' Etat Condense, les Atomes et les Molecules (DSM/DRECAM/LPS-CNRS) UMR9956, 91 - Gif sur Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMULM2E), 91 - Gif-sur-Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMALA2M), 91 - Gif-sur-Yvette (France)

    2007-07-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  9. Reactor cell assembly for use in spectroscopy and microscopy applications

    Science.gov (United States)

    Grindstaff, Quirinus; Stowe, Ashley Clinton; Smyrl, Norm; Powell, Louis; McLane, Sam

    2015-08-04

    The present disclosure provides a reactor cell assembly that utilizes a novel design and that is wholly or partially manufactured from Aluminum, such that reactions involving Hydrogen, for example, including solid-gas reactions and thermal decomposition reactions, are not affected by any degree of Hydrogen outgassing. This reactor cell assembly can be utilized in a wide range of optical and laser spectroscopy applications, as well as optical microscopy applications, including high-temperature and high-pressure applications. The result is that the elucidation of the role of Hydrogen in the reactions studied can be achieved. Various window assemblies can be utilized, such that high temperatures and high pressures can be accommodated and the signals obtained can be optimized.

  10. Developments in acoustic emission for application to nuclear reactor systems

    International Nuclear Information System (INIS)

    Bentley, P.G.

    1982-01-01

    Developments in acoustic emission are summarised as they relate to the principal applications to nuclear reactors, and light water reactor pressure vessels in particular. Improvement in the understanding of acoustic emission has come from materials tests and these confirm the problems in applying the technique for in-service or periodic proof test monitoring of growing fatique cracks. Applications in LMFBR have confirmed that acoustic emission can be applied in the nuclear environment and the detection of stress corrosion cracking in both BWR and LMFBR seems possible. Some information is included on the developing interest in applying the techniques of acoustic emission for leak detection during shop hydro and in-service monitoring. Acoustic emission is also being developed for weld fabrication monitoring and recently introduced pattern recognition techniques are having a significant impact in this application. (author)

  11. High temperature reactors for cogeneration applications

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich (Germany). IEK-6; Allelein, Hans-Josef [Forschungszentrum Juelich (Germany). IEK-6; RWTH Aachen (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik (LRST)

    2016-05-15

    There is a large potential for nuclear energy also in the non-electric heat market. Many industrial sectors have a high demand for process heat and steam at various levels of temperature and pressure to be provided for desalination of seawater, district heating, or chemical processes. The future generation of nuclear plants will be capable to enter the wide field of cogeneration of heat and power (CHP), to reduce waste heat and to increase efficiency. This requires an adjustment to multiple needs of the customers in terms of size and application. All Generation-IV concepts proposed are designed for coolant outlet temperatures above 500 C, which allow applications in the low and medium temperature range. A VHTR would even be able to cover the whole temperature range up to approx. 1 000 C.

  12. Evaluating Russian space nuclear reactor technology for United States applications

    International Nuclear Information System (INIS)

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-01-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch

  13. 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.

  14. 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

  15. Positron annihilation studies on structural materials for nuclear reactors

    International Nuclear Information System (INIS)

    Rajaraman, R.; Amarendra, G.; Sundar, C.S.

    2012-01-01

    Structural steels for nuclear reactors have renewed interest owing to the future advanced fission reactor design with increased burn-up goals as well as for fusion reactor applications. While modified austenitic steels continue to be the main cladding materials for fast breeder reactors, Ferritic/martensitic steels and oxide dispersion strengthened ferritic steels are the candidate materials for future reactors applications in India. Sensitivity and selectivity of positron annihilation spectroscopy to open volume type defects and nano clusters have been extensively utilized in studying reactor materials. We have recently reviewed the application of positron techniques to reactor structural steels. In this talk, we will present successful application of positron annihilation spectroscopy to probe various structural materials such as D9, ferritic/martensitic, oxide dispersion strengthened (ODS) steels and related model alloys, highlighting our recent studies. (author)

  16. SiC/SiC Cladding Materials Properties Handbook

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Singh, Gyanender P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    When a new class of material is considered for a nuclear core structure, the in-pile performance is usually assessed based on multi-physics modeling in coordination with experiments. This report aims to provide data for the mechanical and physical properties and environmental resistance of silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites for use in modeling for their application as accidenttolerant fuel cladding for light water reactors (LWRs). The properties are specific for tube geometry, although many properties can be predicted from planar specimen data. This report presents various properties, including mechanical properties, thermal properties, chemical stability under normal and offnormal operation conditions, hermeticity, and irradiation resistance. Table S.1 summarizes those properties mainly for nuclear-grade SiC/SiC composites fabricated via chemical vapor infiltration (CVI). While most of the important properties are available, this work found that data for the in-pile hydrothermal corrosion resistance of SiC materials and for thermal properties of tube materials are lacking for evaluation of SiC-based cladding for LWR applications.

  17. Effects of corrosion and precipitates on mechanical properties in the ferritic/martensitic steel cladding under ultra-long cycle fast reactor environment at 650 .deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Yong; Lee, Jeong Hyeon; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of); Shin, Sang Hun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This changes chemical compositions of inter-surface and effects on behavior of precipitations. NaCrO{sub 2} which is ternary sodium compound occurs intergranular corrosion resulting in thickness reduction. This change can cause a degradation of mechanical strength of structure material of UCFR. Therefore, we should consider longterm compatibility with sodium and study about life prediction. The research about ferritic/martensitic steel on effects of long term exposure in liquid sodium at 650 .deg. C, 20ppm oxygen includes weight loss of test material (Gr. 92) by corrosion and mechanism about nucleation and growth of precipitates like Laves-phase in bulk. There are many changes such as segregation of component to nucleate precipitates, affecting into microstructural evolution of the steel. Therefore, the thermochemical reaction research to predict behavior about precipitates should be performed. In a specific procedure, the micro-structure and the surface phenomenon of ferritic/martensitic steels (Gr. 92) that are exposed to liquid sodium at 650 .deg. C, 20 ppm oxygen and aged in high pure Argon gas environment to express bulk have been investigated by using scanning electron microscope (SEM) and transmission electron microscope (TEM). At 10 ppm oxygen designed oxygen value for UCFR, there is 107μm thickness reduction for 30 years. Thus, if there is no degradation of mechanical strength caused by aging effect, the tolerance of load of initial cladding should be higher than real load at least 23.6 %. Compared to specimens exposed to Ar-gas environment, Specimen which solutions are leaded into sodium has degradation of strength by reduction of solution hardening.

  18. Naval application of battery optimized reactor integral system

    International Nuclear Information System (INIS)

    Kim, N. H.; Kim, T. W.; Son, H. M.; Suh, K. Y.

    2007-01-01

    Past civilian N.S. Savanna (80 MW t h), Otto-Hahn (38 MW t h) and Mutsu (36 MW t h) experienced stable operations under various sea conditions to prove that the reactors were stable and suitable for ship power source. Russian nuclear icebreakers such as Lenin (90 MW t h x2), Arukuchika (150 MW t h x2) showed stable operations under severe conditions during navigation on the Arctic Sea. These reactor systems, however, should be made even more efficient, compact, safe and long life, because adding support from the land may not be available on the sea. In order to meet these requirements, a compact, simple, safe and innovative integral system named Naval Application Vessel Integral System (NAVIS) is being designed with such novel concepts as a primary liquid metal coolant, a secondary supercritical carbon dioxide (SCO 2 ) coolant, emergency reactor cooling system, safety containment and so on. NAVIS is powered by Battery Optimized Reactor Integral System (BORIS). An ultra-small, ultra-long-life, versatile-purpose, fast-spectrum reactor named BORIS is being developed for a multi-purpose application such as naval power source, electric power generation in remote areas, seawater desalination, and district heating. NAVIS aims to satisfy special environment on the sea with BORIS using the lead (Pb) coolant in the primary system. NAVIS improves the economical efficiency resorting to the SCO 2 Brayton cycle for the secondary system. BORIS is operated by natural circulation of Pb without needing pumps. The reactor power is autonomously controlled by load-following operation without an active reactivity control system, whereas B 4 C based shutdown control rod is equipped for an emergency condition. SCO 2 promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the SCO 2 Brayton

  19. Corrosion Resistant Cladding by YAG Laser Welding in Underwater Environment

    International Nuclear Information System (INIS)

    Tsutomi Kochi; Toshio Kojima; Suemi Hirata; Ichiro Morita; Katsura Ohwaki

    2002-01-01

    It is known that stress-corrosion cracking (SCC) will occur in nickel-base alloys used in Reactor Pressure Vessel (RPV) and Internals of nuclear power plants. A SCC sensitivity has been evaluated by IHI in each part of RPV and Internals. There are several water level instrumentation nozzles installed in domestic BWR RPV. In water level instrumentation nozzles, 182 type nickel-base alloys were used for the welding joint to RPV. It is estimated the SCC potential is high in this joint because of a higher residual stress than the yield strength (about 400 MPa). This report will describe a preventive maintenance method to these nozzles Heat Affected Zone (HAZ) and welds by a corrosion resistant cladding (CRC) by YAG Laser in underwater environment (without draining a reactor water). There are many kinds of countermeasures for SCC, for example, Induction Heating Stress Improvement (IHSI), Mechanical Stress Improvement Process (MSIP) and so on. A YAG laser CRC is one of them. In this technology a laser beam is used for heat source and irradiated through an optical fiber to a base metal and SCC resistant material is used for welding wires. After cladding the HAZ and welds are coated by the corrosion resistant materials so their surfaces are improved. A CRC by gas tungsten arc welding (GTAW) in an air environment had been developed and already applied to a couple of operating plants (16 Nozzles). This method was of course good but it spent much time to perform because of an installation of some water-proof working boxes to make a TIG-weldability environment. CRC by YAG laser welding in underwater environment has superior features comparing to this conventional TIG method as follows. At the viewpoint of underwater environment, (1) an outage term reduction (no drainage water). (2) a radioactive exposure dose reduction for personnel. At that of YAG laser welding, (1) A narrower HAZ. (2) A smaller distortion. (3) A few cladding layers. A YAG laser CRC test in underwater

  20. SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

    Energy Technology Data Exchange (ETDEWEB)

    IJ van Rooyen; WR Lloyd; TL Trowbridge; SR Novascone; KM Wendt; SM Bragg-Sitton

    2013-09-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points

  1. Plasticity of nano-reinforced alloys for the claddings fuels of the fourth generation reactors: experimental and modeling approaches of the effect of microstructural parameters for the models alloys behavior

    International Nuclear Information System (INIS)

    Dade, Mickael

    2015-01-01

    Oxide dispersion strengthened (ODS) ferritic steels are known for their good resistance both to high temperature creep and to swelling under irradiation. They are considered as potential materials for fuel cladding for the next generation of nuclear reactors (Sodium-cooled Fast Reactor). These materials, usually processed by hot extrusion, exhibit a complex microstructure (crystallographic and grain texture, nanometer precipitation, high dislocation density, poly-dispersed grain size), making it a real challenge to establish the microstructure/properties relationships. This thesis has aimed at characterizing and modeling the effect of the different components of the microstructure on the mechanical properties of ferritic Fe-14Cr ODS steels, as well as to improve the understanding of their deformation mechanisms. For this purpose, model materials have been elaborated by hot isostatic pressing and characterized, where the different microstructural parameters have been varied in a controlled manner. Their microstructure have been determined using a set of advanced characterization techniques (SEM-EBSD, TEM, SAXS, EPMA,..). These different materials have been tensile tested over a wide temperature range and creep tested at 650 and 700 C. The results have evidenced the effect of the size and fraction of oxide particles, of the grain size and of the presence of Ti, and have made it possible to model the mechanical behavior. In-situ tensile tests in the SEM, as well as strain field measurements during high temperature testing, have evidenced a transition between a jerky movement of dislocations at low temperature and the high temperature mechanisms, whether intra-granular (dynamic strain ageing, continuous dislocation movement) or inter-granular. At high temperature, severe damage is observed at the grain boundaries. (author) [fr

  2. 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

  3. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    Banerjee, S.; Kamath, H.S.

    2005-01-01

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  4. Liquid metal reactor applications of the CONTAIN code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Gido, R.; Valdez, G.D.; Scholtyssek, W.

    1988-01-01

    The CONTAIN code is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during a severe accident. Included in the phenomena modeled are thermal-hydraulics, radiant and convective heat transfer, aerosol loading and transient response, fission product transport and heating effects, and interactions of sodium and corium with the containment atmosphere and structures. CONTAIN has been used by groups in Japan and West Germany to assess its ability to analyze accident consequences for liquid metal reactor (LMR) plants. In conjunction with this use, collaborative efforts to improve the modeling have been pursued. This paper summarizes the current state of the version of CONTAIN that has been enhanced with extra capabilities for LMR applications. A description of physical models is presented, followed by a review of validation exercises performed with CONTAIN. Some demonstration calculations of an integrated LMR application are presented

  5. Ternary carbide uranium fuels for advanced reactor design applications

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    1999-01-01

    Solid-solution mixed uranium/refractory metal carbides such as the pseudo-ternary carbide, (U, Zr, Nb)C, hold significant promise for advanced reactor design applications because of their high thermal conductivity and high melting point (typically greater than 3200 K). Additionally, because of their thermochemical stability in a hot-hydrogen environment, pseudo-ternary carbides have been investigated for potential space nuclear power and propulsion applications. However, their stability with regard to sodium and improved resistance to attack by water over uranium carbide portends their usefulness as a fuel for advanced terrestrial reactors. An investigation into processing techniques was conducted in order to produce a series of (U, Zr, Nb)C samples for characterization and testing. Samples with densities ranging from 91% to 95% of theoretical density were produced by cold pressing and sintering the mixed constituent carbides at temperatures as high as 2650 K. (author)

  6. Application of vanadium alloys to a fusion reactor blanket

    Energy Technology Data Exchange (ETDEWEB)

    Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

    1984-05-01

    Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

  7. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    International Nuclear Information System (INIS)

    Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

    1977-01-01

    subassemblies in EBR-II and selected encapsulated tests irradiated in GETR. Other comparative tests in thermal reactors indicate that fast flux and thermal flux results are similar. An adequate understanding of FCCI requires integration of in reactor tests and out-of-reactor applied and fundamental studies. To this end a Fuel Cladding Chemical Interaction Program has been established involving several ERDA laboratories and contractors. Other papers to be presented at this international meeting will describe: FCCI work being carried on out-of-reactor but simulating reactor irradiation conditions; studies using all available data sources aimed at illuminating the mechanism and developing models for FCCI; in-reactor and out-of-reactor tests using various techniques and materials whose objectives are to prevent serious FCCI or to mitigate its effect on fuel pin behavior; and the application of FCCI data to lifetime estimates and design criteria

  8. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

    1977-04-01

    subassemblies in EBR-II and selected encapsulated tests irradiated in GETR. Other comparative tests in thermal reactors indicate that fast flux and thermal flux results are similar. An adequate understanding of FCCI requires integration of in reactor tests and out-of-reactor applied and fundamental studies. To this end a Fuel Cladding Chemical Interaction Program has been established involving several ERDA laboratories and contractors. Other papers to be presented at this international meeting will describe: FCCI work being carried on out-of-reactor but simulating reactor irradiation conditions; studies using all available data sources aimed at illuminating the mechanism and developing models for FCCI; in-reactor and out-of-reactor tests using various techniques and materials whose objectives are to prevent serious FCCI or to mitigate its effect on fuel pin behavior; and the application of FCCI data to lifetime estimates and design criteria.

  9. Application of MCNP in the criticality calculation for reactors

    International Nuclear Information System (INIS)

    Zhong Zhaopeng; Shi Gong; Hu Yongming

    2003-01-01

    The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

  10. 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.

  11. Recent improvements in the filtration of corrosion products in high temperature water and application to reactor circuits

    International Nuclear Information System (INIS)

    Darras, R.; Dolle, L.; Chenouard, J.; Laylavoix, F.

    1977-01-01

    The nature and physico-chemical behavior of corrosion products released by structural materials into high temperature water flowing in power reactor circuits have been investigated in test loops and different power plants. The results improve more particularly the knowledge of probable rate constants governing their disappearance through deposition of crud on the fuel cladding. It appears that a considerable limitation of radioactivity transportation in the primary circuit components of pressurized water reactors is in a general way only possible through extraction of the corrosion products by filtration at a rate adequate to minimize the amount of crud deposited in the core. This extraction rate has been estimated; its magnitude implicates a filtration operating on the high temperature water in the primary circuit which allows the necessary high flows. The application of magnetic and electromagnetic so as deep granular graphite bed filters has been studied. The results concerning efficiencies and limiting yields at high temperatures are given. Estimates concerning technological feasibility and corresponding investments are discussed

  12. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  13. Thermodynamics of pellet-cladding interaction

    International Nuclear Information System (INIS)

    Kyoh, Bunkei; Fuji, Kensho

    1987-01-01

    Equilibrium thermodynamic calculations are performed on the U-Zr-Cs-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor (LWR) fuel under pellet-cladding interaction (PCI) failure condition. For this purpose a computer program called SOLGASMIX-PV for the calculation of complex multi-component equilibria is used, and the results of postirradiation examination are interpreted. The analysis of the thermodynamics of the system U-Zr-Cs-I-O indicates that cesium and iodine are assumed to be released from fuel pellet into the fuel-cladding gap as CsI, therefore, the Cs/I ratio in fuel-cladding bonding zone is one. The important condensed phases in this region are UO 2 , U 3 O 8 , Cs 2 U 2 O 7 , Cs 2 U 15 O 46 , ZrO 2 and CsI, and the major gaseous species are CsI, I 2 and I. Under this situation where Cs/I ratio is one, cesium-zirconate is not present. If, however, cesium rich phase is partially present then cesium will be associated with zirconium, possibly as Cs 2 ZrO 3 . (author)

  14. 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.

  15. 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.)

  16. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  17. Standard recommended practice for examination of fuel element cladding including the determination of the mechanical properties

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Guidelines are provided for the post-irradiation examination of fuel cladding and to achieve better correlation and interpretation of the data in the field of radiation effects. The recommended practice is applicable to metal cladding of all types of fuel elements. The tests cited are suitable for determining mechanical properties of the fuel elements cladding. Various ASTM standards and test methods are cited

  18. Kartini Research Reactor prospective studies for neutron scattering application

    International Nuclear Information System (INIS)

    Widarto

    1999-01-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10 7 n/cm 2 s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10 9 n/cm 2 s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  19. Linked Data Reactor: a Framework for Building Reactive Linked Data Applications

    NARCIS (Netherlands)

    Khalili, Ali

    2016-01-01

    This paper presents Linked Data Reactor (LD-Reactor or LD-R) as a framework for developing exible and reusable User Interface components for Linked Data applications. LD-Reactor utilizes Facebook's ReactJS components, Flux architecture and Yahoo's Fluxible framework for isomorphic Web applications.

  20. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  1. Current applications of optimal estimation and control theory to the LOFT reactor plant

    International Nuclear Information System (INIS)

    Feeley, J.J.; Tylee, J.L.

    1980-01-01

    Two advanced estimation and control systems being developed for the LOFT reactor plant are described and evaluated. The advanced protection system, based on a Kalman filter estimator is capable of providing on-line estimates of such critical variables as fuel and cladding temperature, DNBR, and LHGR. The steam generator LQG control system provides stable, closed-loop, zero steady state error control over a wide power range and also provides on-line estimates of certain unmeasureable variables as steam generator power output and cooling capacity for operator information

  2. Current applications of optimal estimation and control theory to the LOFT reactor plant

    International Nuclear Information System (INIS)

    Feeley, J.J.; Tylee, J.L.

    1980-01-01

    Two advanced estimation and control systems being developed for the LOFT reactor plant are described and evaluated. The advanced protection system, based on a Kalman filter estimator is capable of providing on-line estimates of such critical variables as fuel and cladding temperature, DNBR, and LHGR. The steam generator LQG control system provides stable, closed-loop, zero steady state error control over a wide power range and also provides on-line estimates of certain unmeasureable variables as steam generator power output and cooling capacity for operator information. 12 refs

  3. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Magny, E. de; Berte, M.

    1997-01-01

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

  4. Assessment of very high temperature reactors in process applications

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.; Gambill, W.R.

    1976-01-01

    In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

  5. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  6. Reactor noise diagnostics based on multivariate autoregressive modeling: Application to LOFT [Loss-of-Fluid-Test] reactor process noise

    International Nuclear Information System (INIS)

    Gloeckler, O.; Upadhyaya, B.R.

    1987-01-01

    Multivariate noise analysis of power reactor operating signals is useful for plant diagnostics, for isolating process and sensor anomalies, and for automated plant monitoring. In order to develop a reliable procedure, the previously established techniques for empirical modeling of fluctuation signals in power reactors have been improved. Application of the complete algorithm to operational data from the Loss-of-Fluid-Test (LOFT) Reactor showed that earlier conjectures (based on physical modeling) regarding the perturbation sources in a Pressurized Water Reactor (PWR) affecting coolant temperature and neutron power fluctuations can be systematically explained. This advanced methodology has important implication regarding plant diagnostics, and system or sensor anomaly isolation. 6 refs., 24 figs

  7. Gas reactor international cooperative program. HTR-synfuel application assessment

    International Nuclear Information System (INIS)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H 2 +CO 2 ) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000 0 F steam is generated at the industrial user sites. The products of methanation (CH 4 + H 2 O) are piped back to the reformer at the central station HTR

  8. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  9. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding.

  10. Utilization of research reactors in universities and their medical applications

    International Nuclear Information System (INIS)

    Kanda, Keiji.

    1983-01-01

    In Japan, five research reactors and a critical assembly are operated by the universities. They are opened to all university researchers, the system of which is financially supported by the Ministry of Education, Culture and Science of the Japanese government. Usually KUR is operated eight cycles per year. One cycle consists of the following four week operation: 1. Mainly for researchers from other universities; 2. Mainly for researchers in the institute; 3. Mainly for beam experiment; 4. Sort time experiment. In the weeks of 1 ∼ 3 the KUR is operated continously from Tuesday morning to Friday evening. The experiment include studies on physics, chemistry, biology, medicine, engineering etc. Recently the medical application of research reactors has become popular in Japan. The new technique of the boron neutron capture thereby has been successfully applied to brain tumors and will be to melanoma (skin cancer) in near future. (author)

  11. Artificial Intelligence and its Reasonable Application Scenario to Reactor Operation

    International Nuclear Information System (INIS)

    Im, Ki Hong; Suh, Yong-Suk; Park, Cheol; Lim, In-Cheol

    2017-01-01

    This paper presents brief but reasonable scenarios for applying AI or machine learning technologies to research reactor from various perspectives. Two less safety critical scenarios for applying AI to reactor operation are introduced in this study. However, the AI assistant will not only be an assistant but it will also be an operator in the future. What is required is big operation data which can represent all the cases requiring operation decision, including normal operation and accident data as well, and enough time to train and fix the AI system with this data. We can predict AI study in this area can begin with a mild and safe application. But in the near future, this technology could be used to handle or automate more severe operations.

  12. 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)

  13. Cladding failure by local plastic instability

    International Nuclear Information System (INIS)

    Kramer, J.M.; Deitrich, L.W.

    1977-01-01

    Cladding failure is one of the major considerations in analysis of fast-reactor fuel pin behavior during hypothetical accident transients since time, location and nature of failure govern the early post-failure material motion and reactivity feedback. Out-of-Pile transient burst tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture. To investigate the details of cladding bulging, a perturbation analysis of the equations governing the large deformation of a cylindrical shell has been developed. The overall deformation history is assumed to consist of a small perturbation epsilon of the radial displacement superimposed on large axisymmetric displacements. Computations have been carried out using high temperature properties of stainless steel in conjunction with various constitutive theories, including a generalization of the Endochronic Theory of Plasticity in which both time-independent and time-dependent plastic strains are modeled. Although the results of the calculations are all qualitatively similar, it appears that modeling of both time-independent and time-dependent plastic strains is necessary to interpret the transient burst test results. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or non-symmetric cooling. (Auth.)

  14. The characterization of activities associated with irradiated fuel element claddings

    International Nuclear Information System (INIS)

    Jenkins, I.L.; Bolus, D.J.; Glover, K.M.; Haynes, J.W.; Mapper, D.; Marwick, A.D.; Waterman, M.J.

    1982-01-01

    The object of the present work was to characterise the natures and amounts of the various α and βγ activities associated with cladding hulls. The claddings studied were stainless steel from a Fast Reactor and from an Advanced Gas Reactor and Zircaloy from a Boiling Water Reactor, from a Pressurized Water Reactor and from a Steam Generating Heavy Water Reactor. The hulls were examined by the following methods: alpha spectrometry to identify and quantify the α emitters and to estimate their depths of penetration, partial and complete dissolution of hulls followed by gross α counting, α spectrometry and γ spectrometry, fission track autoradiography to determine the distribution of fissile material associated with hulls, neutron activation to determine the total fissile content of the hulls, chemical separations followed by β counting and chemical treatment with various reagents to examine the ease of decontamination

  15. 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)

  16. 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

  17. The possibilities of application of experimental Kfk results from BR2 on SNR designs

    International Nuclear Information System (INIS)

    Karsten, G.; Elbel, K.; Dienst, W.; Schaefer, L.

    1978-01-01

    A review is given of the relevant results of the technological application for the SNR300 reactor, since the BR2 reactor has been used as a test facility for the material development. Special emphasis has been laid on the fuel pin behavior under the aspect of chemical and mechanical fuel-clad interaction and on the specification of the cladding in terms of high temperature mechanical behavior in the SNR 300 reactor. A systematic analysis of urgent research topics in BR2 test facility reactor is presented. (A.F.)

  18. Application of damage function analysis to reactor coolant circuits

    International Nuclear Information System (INIS)

    MacDonald, D.D.

    2002-01-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  19. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  20. Voltage-Dependent Charge Storage in Cladded Zn0.56Cd0.44Se Quantum Dot MOS Capacitors for Multibit Memory Applications

    Science.gov (United States)

    Khan, J.; Lingalugari, M.; Al-Amoody, F.; Jain, F.

    2013-11-01

    As conventional memories approach scaling limitations, new storage methods must be utilized to increase Si yield and produce higher on-chip memory density. Use of II-VI Zn0.56Cd0.44Se quantum dots (QDs) is compatible with epitaxial gate insulators such as ZnS-ZnMgS. Voltage-dependent charging effects in cladded Zn0.56Cd0.44Se QDs are presented in a conventional metal-oxide-semiconductor capacitor structure. Charge storage capabilities in Si and ZnMgS QDs have been reported by various researchers; this work is focused on II-VI material Zn0.56Cd0.44Se QDs nucleated using photoassisted microwave plasma metalorganic chemical vapor deposition. Using capacitance-voltage hysteresis characterization, the multistep charging and discharging capabilities of the QDs at room temperature are presented. Three charging states are presented within a 10 V charging voltage range. These characteristics exemplify discrete charge states in the QD layer, perfect for multibit, QD-functionalized high-density memory applications. Multiple charge states with low operating voltage provide device characteristics that can be used for multibit storage by allowing varying charges to be stored in a QD layer based on the applied "write" voltage.

  1. Report on the Survey of the Design Review of New Reactor Applications. Volume 3: Reactor

    International Nuclear Information System (INIS)

    Downey, Steven; Monninger, John; Nevalainen, Janne; Lorin, Aurelie; ); Webster, Philip; Joyer, Philippe; Kawamura, Tomonori; Lankin, Mikhail; Kubanyi, Jozef; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Kang, Kyungmin; Kim, Walter

    2016-01-01

    At the tenth meeting of the CNRA Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the Licensing Process Survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This document, which is the third report on the results of the Design Phase Survey, focuses on the Reactor. The Reactor category includes the following technical topics: fuel system design, reactor internals and core support, nuclear design and core nuclear performance, thermal and hydraulic design, reactor materials, and functional design of reactivity control system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - Design review strategies most commonly used to confirm that the regulatory requirements have been met include document review and independent verification of calculations, computer codes, or models used to describe the design and performance of the core and the fuel. - It is common to consider operating experience and

  2. 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

  3. Future possibilities of SUSEN technologies for R&D of nuclear fuel cladding

    International Nuclear Information System (INIS)

    Mikloš, M.

    2015-01-01

    R&D possibilities with nuclear fuel cladding were discussed in this paper. The availability of 10 MWT reactor with BWR and PWR loops having chemistry control was described. Activity transport and fuel cladding corrosion can be investigated in this facility including PIE. The facility has hot cells and the laboratory is expected to start in 2017

  4. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    International Nuclear Information System (INIS)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented

  5. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    International Nuclear Information System (INIS)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures

  6. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

    Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

  7. Spent fuel cladding containment credit test

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1983-01-01

    As an initial step in addressing the effectiveness of breached cladding as a barrier to radionuclide release from the repository during the post-containment period, preliminary scoping tests have been initiated which compare radionuclide releases from spent fuel specimens with artificially induced cladding defects of various severities. The artificially induced defects are all more severe than the typical in-reactor type breaches which are expected to be the principal type of breach entering the repository for terminal storage. These preliminary scoping tests being conducted by Westinghouse Hanford Company for the Lawrence Livermore National Laboratory Waste Package Development Program in support of the Tuff repository project at the Nevada Test Site are described. Also included in this presentation are selected initial results from these tests. 22 figures

  8. Creep collapse of TAPS fuel cladding

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Anand, A.K.

    1975-01-01

    Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported (internally) at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : (1) inelastic collapse of perfectly circular tubes (2) plastic instability in oval tubes (3) effect of creep on ovality. Creep is considered to be a non-linear combination of the following : (a) thermal creep (b) intresenic creep (c) stress aided radiation enhanced (d) stress free growth (4) Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. (author)

  9. 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

  10. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  11. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

    1987-12-01

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

  12. Space nuclear reactor shields for manned and unmanned applications

    International Nuclear Information System (INIS)

    McKissock, B.I.; Bloomfield, H.S.

    1990-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

  13. Space nuclear reactor shields for manned and unmanned applications

    International Nuclear Information System (INIS)

    Mckissock, B.I.; Bloomfield, H.S.

    1989-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

  14. APPLICATION OF MEMBRANE SORPTION REACTOR TECHNOLOGY FOR LRW MANAGEMENT

    International Nuclear Information System (INIS)

    Glagolenko, Yuri; Dzekun, Evgeny; Myasoedovg, Boris; Gelis, Vladimir; Kozlitin, Evgeny; Milyutin, Vitaly; Trusov, Lev; Rengel, Mike; Mackay, Stewart M.; Johnson, Michael E.

    2003-01-01

    A new membrane-sorption technology has been recently developed and industrially implemented in Russia for the treatment of the Liquid (Low-Level) Radioactive Waste (LRW). The first step of the technology is a precipitation of the radionuclides and/or their adsorption onto sorbents of small particle size. The second step is filtration of the precipitate/sorbent through the metal-ceramic membrane, Trumem.. The unique feature of the technology is a Membrane-Sorption Reactor (MSR), in which the precipitation / sorption and the filtration of the radionuclides occur simultaneously, in one stage. This results in high efficiency, high productivity and compactness of the equipment, which are the obvious advantages of the developed technology. Two types of MSR based on Flat Membranes device and Centrifugal Membrane device were developed. The advantages and disadvantages of application of each type of the reactors are discussed. The MSR technology has been extensively tested and efficiently implemented at ''Mayak '' nuclear facility near Chelyabinsk, Russia as well as at other Russian sites. The results of this and other applications of the MSR technology at the different Russian nuclear facilities are discussed. The results of the first industrial applications of the MSR technology for radioactive waste treatment in Russia and analysis of the available information about LRW accumulated in other countries imply that this technology can be successfully used for the Low Level Radioactive Waste treatment in the USA and in other nuclear countries

  15. Some applications of capacitance technology in nuclear reactor components inspections

    International Nuclear Information System (INIS)

    Walton, H.

    1985-01-01

    The paper considers application of a capacitance measuring system that has overcome many of the original contraints, such as sensitivity to cable length, induced electric field and high acoustic noise, and illustrates the ease of use with examples of proven capability in severe environments of high temperature or high radiation. The Capacitance Displacement Transducer (CDT) measuring principle was originally developed as a working technique during the early years of full-scale, on-load refuelling trials performed in the Windscale Civil Advanced Gas-Cooled Reactor (CAGR) test rig where it was necessary to measure the vibrational behaviour of fuel components in simulated reactor conditions. At that time, 1968-1969, no instrumentation existed that would measure displacement in the range 0 to 100 mms to an accuracy of 25x10 -3 mms, without physical contact, at temperatures of 600 0 C in high velocity gas, in high acoustic noise fields of 150 db's over cable lengths approaching 100 metres. The principles incorporated in the CDT overcome all these problems. The advantages inherent in this system have been extended to metrology applications in more recent years by the further development of the electronics to enable linear displacement measurement to be obtained between two capacitance plates whose separation varies, either by plate movement or by surface irregularity. This principle has been used to good effect in novel applications associated with the inspection of nominally inaccessible internal tube surfaces

  16. Proceedings of the Conference on research reactors application in Yugoslavia

    International Nuclear Information System (INIS)

    1978-05-01

    The Conference on research reactors operation was organised on the occasion of 20 anniversary of the RB zero power reactor start-up. The presentations showed that research reactors in Yugoslavia, RB, RA and TRIGA had an important role in development of nuclear sciences and technology in Yugoslavia. The reactors were applied in non-destructive testing of materials and fuel elements, development of reactor noise techniques, safety analyses, reactor control methods, neutron activation analysis, neutron radiography, dosimetry, isotope production, etc [sr

  17. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  18. New applications of neutron noise theory in power reactor physics

    International Nuclear Information System (INIS)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  19. Application of the SHARP Toolkit to Sodium-Cooled Fast Reactor Challenge Problems

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yu, Y. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2017-09-30

    The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit is under development by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of the U.S. Department of Energy, Office of Nuclear Energy. To better understand and exploit the benefits of advanced modeling simulations, the NEAMS Campaign initiated the “Sodium-Cooled Fast Reactor (SFR) Challenge Problems” task, which include the assessment of hot channel factors (HCFs) and the demonstration of zooming capability using the SHARP toolkit. If both challenge problems are resolved through advanced modeling and simulation using the SHARP toolkit, the economic competitiveness of a SFR can be significantly improved. The efforts in the first year of this project focused on the development of computational models, meshes, and coupling procedures for multi-physics calculations using the neutronics (PROTEUS) and thermal-hydraulic (Nek5000) components of the SHARP toolkit, as well as demonstration of the HCF calculation capability for the 100 MWe Advanced Fast Reactor (AFR-100) design. Testing the feasibility of the SHARP zooming capability is planned in FY 2018. The HCFs developed for the earlier SFRs (FFTF, CRBR, and EBR-II) were reviewed, and a subset of these were identified as potential candidates for reduction or elimination through high-fidelity simulations. A one-way offline coupling method was used to evaluate the HCFs where the neutronics solver PROTEUS computes the power profile based on an assumed temperature, and the computational fluid dynamics solver Nek5000 evaluates the peak temperatures using the neutronics power profile. If the initial temperature profile used in the neutronics calculation is reasonably accurate, the one-way offline method is valid because the neutronics power profile has weak dependence on small temperature variation. In order to get more precise results, the proper temperature profile for initial neutronics calculations was obtained from the

  20. Application of the SPEEDI system to the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Chino, Masamichi; Ishikawa, Hirohiko; Yamazawa, Hiromi; Moriuchi, Shigeru

    1986-10-01

    The SPEEDI system is a computational code system to predict the radiological dose due to the plume released in a nuclear accident in Japan. This paper describes the SPEEDI's application to the Chernobyl reactor accident for the estimation of the movement of plume and the release rate of radioactive nuclides into the environment. The predicted results on the movement of plume agreed well with the monitoring data in Europe. The estimated results on the release rate showed that half of the noble gas inventory, about 5 % of the iodine inventory and about 3 % of the cesium inventory are released into the environment within 24 hours. (author)