WorldWideScience

Sample records for fuel assembly clads

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

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R. B.; Woods, K. N. [Siemens Nuclear Power Corp., Richland, WA (United States)

    1997-04-01

    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.

  2. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Gyun; Kim, Young Il

    2006-12-15

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006.

  3. EPRI fuel cladding integrity program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, R. [Electric Power Research Institute, Palo Alto, CA (United States)

    1997-01-01

    The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R&D includes fuel and cladding. In this paper, only R&D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response.

  4. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  5. Fuel pin cladding

    Science.gov (United States)

    Vaidyanathan, S.; Adamson, M.G.

    1986-01-28

    Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

  6. Fuel pin cladding

    Science.gov (United States)

    Vaidyanathan, Swaminathan; Adamson, Martyn G.

    1986-01-01

    An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

  7. Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

    Science.gov (United States)

    Rohal, R. G.; Tambling, T. N.; Smith, R. L.

    1973-01-01

    Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

  8. Supplemental information for a notice of construction 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 ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). 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 to be used 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, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  9. 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.

  10. Water-moderated reactor fuel cladding reliability study

    OpenAIRE

    Бакутяк, Елена Викторовна; Пелых, Сергей Николаевич

    2014-01-01

    Considering the fuel element, averaged by fuel assembly (FA) of water-moderated reactor with the power of 1000 MW (VVER-1000), the number of fuel elements with the greatest cladding failure probability after 4 operation years at Khmelnitsky NPP-2 (KNPP-2) is found. This will allow to calculate the fuel cladding failure probability and determine the most likely cladding damages, which will enable to improve the performance and economic indexes of VVER.The novelty of the paper lies in calculati...

  11. High energy X-ray diffraction measurement of residual stresses in a monolithic aluminum clad uranium-10 wt% molybdenum fuel plate assembly

    Science.gov (United States)

    Brown, D. W.; Okuniewski, M. A.; Almer, J. D.; Balogh, L.; Clausen, B.; Okasinski, J. S.; Rabin, B. H.

    2013-10-01

    Residual stresses are expected in monolithic, aluminum clad uranium 10 wt% molybdenum (U-10Mo) nuclear fuel plates because of the large mismatch in thermal expansion between the two bonded materials. The full residual stress tensor of the U-10Mo foil in a fuel plate assembly was mapped with 0.1 mm resolution using high-energy (86 keV) X-ray diffraction. The in-plane stresses in the U-10Mo foil are strongly compressive, roughly -250 MPa in the longitudinal direction and -140 MPa in the transverse direction near the center of the fuel foil. The normal component of the stress is weakly compressive near the center of the foil and tensile near the corner. The disparity in the residual stress between the two in-plane directions far from the edges and the tensile normal stress suggest that plastic deformation in the aluminum cladding during fabrication by hot isostatic pressing also contributes to the residual stress field. A tensile in-plane residual stress is presumed to be present in the aluminum cladding to balance the large in-plane compressive stresses in the U-10Mo fuel foil, but cannot be directly measured with the current technique due to large grain size.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

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

  13. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jaramillo, Roger A [ORNL; Hendrich, WILLIAM R [ORNL; Packan, Nicolas H [ORNL

    2007-03-01

    A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were

  14. A method for limitation of probability of accumulation of fuel elements claddings damage in WWER

    OpenAIRE

    Sergey N. Pelykh; Mark V. Nikolsky; S. D. Ryabchikov

    2014-01-01

    The aim is to reduce the probability of accumulation of fuel elements claddings damage by developing a method to control the properties of the fuel elements on stages of design and operation of WWER. An averaged over the fuel assembly WWER-1000 fuel element is considered. The probability of depressurization of fuel elements claddings is found. The ability to predict the reliability of claddings by controlling the factors that determine the properties of the fuel elements is proved. The expedi...

  15. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  16. Fuel assembly reconstitution

    Energy Technology Data Exchange (ETDEWEB)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos, E-mail: mongeor@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil)

    2009-07-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  17. Nuclear fuel elements having a composite cladding

    Science.gov (United States)

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  18. Thermal Analysis of a TREAT Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, Dionissios [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, Arthur E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  19. Double-clad nuclear fuel safety rod

    Science.gov (United States)

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  20. 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.

  1. Clad thickness variation N-Reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, E.A.

    1966-05-12

    The current specifications for the cladding on {open_quotes}N{close_quotes} fuels were established early in the course of process development and were predicted on several basic considerations. Among these were: (a) a desire to provide an adequate safety factor in cladding thickness to insure against corrosion penetration and rupture from uranium swelling stresses; (b) an apprehension that the striations in the zircaloy cladding of the U/zircaloy interface and on the exterior surface might serve as stress-raisers, leading to untimely failures of the jacket; and (c) then existing process capability - the need to maintain a specified ratio between zircaloy and uranium in the billet assembly to effect satisfactory coextrusion. It now appears appropriate to review these specifications in an effort to determine whether some of them may be revised, with attendant gains in economy and/or operating smoothness.

  2. CLAD CARBIDE NUCLEAR FUEL, THERMIONIC POWER, MODULES.

    Science.gov (United States)

    The general objective is to evaluate a clad carbide emitter, thermionic power module which simulates nuclear reactor installation, design, and...performance. The module is an assembly of two series-connected converters with a single common cesium reservoir. The program goal is 500 hours

  3. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Gilbert, E.R.

    1983-09-01

    This report summarizes the technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. In addition, dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor (PWR) fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved {similar_to}5,000 fuel rods, and {similar_to}600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570{sup 0}C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at {similar_to}70{sup 0}C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the United States. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380{sup 0}C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400{sup 0}C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved.

  4. PFR fuel cladding transient test results and analysis

    Science.gov (United States)

    Cannon, N. S.; Hunter, C. W.; Kear, K. L.; Wood, M. H.

    1986-05-01

    Fuel Cladding Transient Tests (FCTT) were performed on M316 cladding specimens obtained from mixed-oxide fuel pins irradiated in the Prototype Fast Reactor (PFR) to burnups of 4 and 9 atom percent. In these tests, specimens of fuel cladding were pressurized and heated until failure occurred. Samples of cladding from PFR fuel pins exhibited generally greater strength and ductility than specimens from Experimental Breeder Reactor-II (EBR-II) mixed-oxide fuel pins tested under similar conditions. Apparently, the PFR cladding properties were not degraded by a fuel adjacency effect (FAE) observed in fuel pin cladding from EBR-II irradiations. A recently developed model of grain boundary cavity growth was used to predict the results of the tests conducted on PFR cladding. It was found that the predicted failure temperatures for the relevant internal pressures were in good agreement with experimental failure temperatures.

  5. 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)

  6. Fuel clad chemical interactions in fast reactor MOX fuels

    Science.gov (United States)

    Viswanathan, R.

    2014-01-01

    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 ṡ [B/(at.% fission)] ṡ (T/K-705) ṡ [(O/M)i-1.935]} + 20.5) for (O/M)i ⩽ 1.98. A new model is proposed for (O/M)i ⩾ 1.98: d/μm = [B/(at.% fission)] ṡ (T/K-800)0.5 ṡ [(O/M)i-1.94] ṡ [P/(W cm-1)]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)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.

  7. Results from studies of surface deposits on the claddings of fuel rods used in RBMK-1000 reactors

    Science.gov (United States)

    Smirnova, I. M.; Markov, D. V.

    2010-07-01

    The results of studies on analyzing the element composition of deposits on the cladding surfaces of fuel rods used in a fuel assembly at the Leningrad nuclear power station are presented. The distribution of elements in deposits over the fuel rod height is analyzed, and the zones of their concentration are revealed. It is shown that deposits of copper penetrating into cracks in the surface layer of zirconium oxide introduce an essential contribution in the development of nodular corrosion of fuel rod claddings.

  8. 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.)

  9. Accident-tolerant oxide fuel and cladding

    Energy Technology Data Exchange (ETDEWEB)

    Mariani, Robert D.

    2017-05-30

    Systems and methods for accident tolerant oxide fuel. One or more disks can be placed between fuel pellets comprising UO.sub.2, wherein such disks possess a higher thermal conductivity material than that of the UO.sub.2 to provide enhanced heat rejection thereof. Additionally, a cladding coating comprising zircaloy coated with a material that provides stability and high melting capability can be provided. The pellets can be configured as annular pellets having an annulus filled with the higher thermal conductivity material. The material coating the zircaloy can be, for example, Zr.sub.5Si.sub.4 or another silicide such as, for example, a Zr-Silicide that limits corrosion. The aforementioned higher thermal conductivity material can be, for example, Si, Zr.sub.xSi.sub.y, Zr, or Al.sub.2O.sub.3.

  10. Nuclear reactor fuel element with vanadium getter on cladding

    Science.gov (United States)

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  11. Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel

    Science.gov (United States)

    Carmack, William J.

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors and provide a number of advantages over other fuel types considering their fabricability, performance, recyclability, and safety. Resistance to cladding "breach" and subsequent release of fission products and fuel constituents to the nuclear power plant primary coolant system is a key performance parameter for a nuclear fuel system. In metallic fuel, FCCI weakens the cladding, especially at high power-high temperature operation, contributing to fuel pin breach. Empirical relationships for FCCI have been developed from a large body of data collected from in-pile (EBR-II) and out-of-pile experiments [1]. However, these relationships are unreliable in predicting FCCI outside the range of EBR-II experimental data. This dissertation examines new FCCI data extracted from the MFF-series of prototypic length metallic fuel irradiations performed in the Fast Flux Test Facility (FFTF). The fuel in these assemblies operated a temperature and burnup conditions similar to that in EBR-II but with axial fuel height three times longer than EBR-II experiments. Comparing FCCI formation data from FFTF and EBR-II provides new insight into FCCI formation kinetics. A model is developed combining both production and diffusion of lanthanides to the fuel-cladding interface and subsequent reaction with the cladding. The model allows these phenomena to be influenced by fuel burnup (lanthanide concentrations) and operating temperature. Parameters in the model are adjusted to reproduce measured FCCI layer thicknesses from EBR-II and FFTF. The model predicts that, under appropriate conditions, rate of FCCI formation can be controlled by either fission product transport or by the reaction rate of the interaction species at the fuel-cladding interface. This dissertation will help forward the design of metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full

  12. FUEL ASSEMBLY SHAKER TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  13. Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

    Science.gov (United States)

    Li, Bo-Shiuan

    Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission

  14. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

  15. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

  16. The Welding Process of the Small In-pile Testing Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The small in-pile testing fuel assembly is designed for high performance fuel assembly study. It has two parts of which are four fuel element with double layer cladding and a detect system for measurement of testing pressure and temperature. The fuel element is composed of UO2 pellets, the stainless steel cladding and end caps. The detect system is direct contact with the fuel element by electron beam welding. In the fabrication of the assembly, some special welding technologies are

  17. Characterization of Hydrogen Content in ZIRCALOY-4 Nuclear Fuel Cladding

    Science.gov (United States)

    Pfeif, E. A.; Lasseigne, A. N.; Krzywosz, K.; Mader, E. V.; Mishra, B.; Olson, D. L.

    2010-02-01

    Assessment of hydrogen uptake of underwater nuclear fuel clad and component materials will enable improved monitoring of fuel health. Zirconium alloys are used in nuclear reactors as fuel cladding, fuel channels, guide tubes and spacer grids, and are available for inspection in spent fuel pools. With increasing reactor exposure zirconium alloys experience hydrogen ingress due to neutron interactions and water-side corrosion that is not easily quantified without destructive hot cell examination. Contact and non-contact nondestructive techniques, using Seebeck coefficient measurements and low frequency impedance spectroscopy, to assess the hydrogen content and hydride formation within zircaloy 4 material that are submerged to simulate spent fuel pools are presented.

  18. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  19. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  20. Intercode Advanced Fuels and Cladding Comparison Using BISON, FRAPCON, and FEMAXI Fuel Performance Codes

    Science.gov (United States)

    Rice, Aaren

    As part of the Department of Energy's Accident Tolerant Fuels (ATF) campaign, new cladding designs and fuel types are being studied in order to help make nuclear energy a safer and more affordable source for power. This study focuses on the implementation and analysis of the SiC cladding and UN, UC, and U3Si2 fuels into three specific nuclear fuel performance codes: BISON, FRAPCON, and FEMAXI. These fuels boast a higher thermal conductivity and uranium density than traditional UO2 fuel which could help lead to longer times in a reactor environment. The SiC cladding has been studied for its reduced production of hydrogen gas during an accident scenario, however the SiC cladding is a known brittle and unyielding material that may fracture during PCMI (Pellet Cladding Mechanical Interaction). This work focuses on steady-state operation with advanced fuel and cladding combinations. By implementing and performing analysis work with these materials, it is possible to better understand some of the mechanical interactions that could be seen as limiting factors. In addition to the analysis of the materials themselves, a further analysis is done on the effects of using a fuel creep model in combination with the SiC cladding. While fuel creep is commonly ignored in the traditional UO2 fuel and Zircaloy cladding systems, fuel creep can be a significant factor in PCMI with SiC.

  1. 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.

  2. Composite nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Dollard, W.J.; Ferrari, H.M.

    1982-04-27

    An open lattice elongated nuclear fuel assembly including small diameter fuel rods disposed in an array spaced a selected distance above an array of larger diameter fuel rods for use in a nuclear reactor having liquid coolant flowing in an upward direction. Plenums are preferably provided in the upper portion of the upper smaller diameter fuel rods and in the lower portion of the lower larger diameter fuel rods. Lattice grid structures provide lateral support for the fuel rods and preferably the lowest grid about the upper rods is directly and rigidly affixed to the highest grid about the lower rods.

  3. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  4. Probabilistic Failure Analysis for Wound Composite Ceramic Cladding Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hemrick, James Gordon [ORNL; Lara-Curzio, Edgar [ORNL

    2013-01-01

    Advanced ceramic matrix composites based on silicon carbide (SiC) are being considered as candidate material systems for nuclear fuel cladding in light water reactors. The SiC composite structure is considered due to its assumed exceptional performance under accident scenarios, where its excellent high-temperature strength and slow reaction kinetics with steam and associated mitigated hydrogen production are desirable. The specific structures of interest consist of a monolithic SiC cylinder surrounded by interphase-coated SiC woven fibers in a tubular form and infiltrated with SiC. Additional SiC coatings on the outermost surface of the assembly are also being considered to prevent hydrothermal corrosion of the fibrous structure. The inner monolithic cylinder is expected to provide a hermetic seal to contain fission products under normal conditions. While this approach offers the promise of higher burn-up rates and safer behavior in the case of LOCA events, the reliability of such structures must be demonstrated in advance. Therefore, a probability failure analysis study was performed of such monolithic-composite hybrid structures to determine the feasibility of these design concepts. This analysis will be used to predict the future performance of candidate systems in an effort to determine the feasibility of these design concepts and to make future recommendations regarding materials selection.

  5. Modeling the mechanical behaviour of CANDU fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Holt, R.A. [Queen' s Univ., Dept. of Mechanical Engineering, Kingston, Ontario (Canada)

    2003-07-01

    Models for the mechanical behaviour of fuel cladding were developed in the period 1973-1983 by staff at AECL CRNL. The models for the mechanical properties of fuel cladding during normal operation were a by-product of programs during the period 1970-1975 to understand the origin of fuel-cladding defects caused by power ramps at Douglas Point and Pickering A. Models for accident conditions were, initially, based heavily on mechanical properties data generated by McGill University and Westinghouse Canada under contract to AECL in the late 1960's and early 1970's and attempts to interpret the data in terms of the underlying deformation mechanisms. The model for normal operating conditions was embodied in the ELESTRES/ELESIM series of codes, and the models for accident conditions were embodied in NIRVANA. (author)

  6. Characterization of Fuel-Cladding Bond Strength Using Laser Shock

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; David L. Cottle; Barry H. Rabin

    2014-04-01

    This paper describes new laser-based capabilities for characterization of fuel-cladding bond strength in nuclear fuels, and presents preliminary results obtained from studies on as-fabricated monolithic fuel consisting of uranium-10 wt.% molybdenum alloys clad in 6061 aluminum by hot isostatic pressing. Two complementary experimental methods are employed, laser-shock testing and laser-ultrasonic imaging. Measurements are spatially localized, non-contacting and require minimum specimen preparation, and are therefore ideally suited for applications involving radioactive materials, including irradiated materials. The theoretical principles and experimental approaches employed in characterization of nuclear fuel plates are described. The ability to measure layer thicknesses, elastic properties of the constituents, and the location and nature of laser-shock induced debonds is demonstrated, and preliminary bond strength measurement results are discussed.

  7. Impact of thicker cladding on the nuclear parameters of the NPP Krsko fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kromar, Marjan, E-mail: marjan.kromar@ijs.s [Jozef Stefan Institute, Reactor Physics Department, Jamova 39, 1001 Ljubljana (Slovenia); Kurincic, Bojan [Nuclear Power Plant Krsko, Engineering Division, Nuclear Fuel and Reactor Core, Vrbina 12, 8270 Krsko (Slovenia)

    2011-04-15

    To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krsko that uses 16 x 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.

  8. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

  9. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  10. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Science.gov (United States)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

  11. Novel Accident-Tolerant Fuel Meat and Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas release and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.

  12. Microbial Biofilm Growth on Irradiated, Spent Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    S.M. Frank

    2009-02-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.

  13. Material Selection for Accident Tolerant Fuel Cladding

    Science.gov (United States)

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as >100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥1473 K (1200 °C) for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases, and FeCrAl alloys. Recently reported low-mass losses for Mo in steam at 1073 K (800 °C) could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1473 K (1200 °C) in steam and significant TiO2, and therefore, Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1748 K (1475 °C), while reducing its Cr content to minimize susceptibility to irradiation-assisted α' formation. The composition effects and critical limits to retaining protective scale formation at >1673 K (1400 °C) are still being evaluated.

  14. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  15. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

  16. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  17. 78 FR 3853 - Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an Independent Spent Fuel...

    Science.gov (United States)

    2013-01-17

    ... COMMISSION 10 CFR Parts 71 and 72 Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an... several key areas, such as: retrievability, cladding integrity, and safe handling of spent fuel... potential policy issues and requirements related to retrievability, cladding integrity, and safe handling...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-01

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

  19. Most advanced HTP fuel assembly design for EPR

    Energy Technology Data Exchange (ETDEWEB)

    Francillon, Eric [AREVA - Framatome ANP, 10 rue Juliette Recamier - 69456 Lyon Cedex 06 (France); Kiehlmann, Horst-Dieter [AREVA - Framatome ANP GmbH, P.O. Box 3220, 91050 Erlangen (Germany)

    2006-07-01

    End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-09-01

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

  2. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-12-14

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows for

  3. Fuel cell sub-assembly

    Science.gov (United States)

    Chi, Chang V.

    1983-01-01

    A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

  4. Experimental Setup for Reflood Quench of Accident Tolerant Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan; Lee, Kwan Geun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The concept of accident tolerant fuel (ATF) is a solution to suppress the hydrogen generation in loss of coolant accident (LOCA) situation without safety injection, which was the critical incident in the severe accident in the Fukushima. The changes in fuel and cladding materials may cause a significant difference in reactor performance in long term operation. Properties in terms of material science and engineering have been tested and showed promising results. However, numerous tests are still required to ensure the design performance and safety. Thermal hydraulic tests including boiling and quenching are partly confirmed, but not yet complete. We have been establishing the experimental setup to confirm the properties in the terms of thermal hydraulics. Design considerations and preliminary tests are introduced in this paper. An experimental setup to test thermal hydraulic characteristics of new ATF claddings are established and tested. The W heater set inside the cladding is working properly, exceeding 690 W/m linear power with thermocouples and insulating ceramic sheaths inside. The coolant injection control was also working in good conditions. The setup is about to complete and going to simulate quenching behavior of the ATF in the LOCA situation.

  5. ZrC COATING ON FUEL ELEMENT CLADDING ZIRCALOY-2

    Directory of Open Access Journals (Sweden)

    Etty Mutiara

    2017-02-01

    Full Text Available ZrC COATING ON FUEL ELEMENT ZIRCALOY-2 CLADDING. The intensive researchs on high discharge burn-up of Light Water Reactor (LWR fuel element were performed due to the extension of fuel element’s utility life. One of these researches was allowing for alteration of the existing zirconium-based clad system through coating. This technique is supposed to improve the corrosion resistance of cladding without changing the dimension of fuel cladding. In current research, the ZrC film was coated on the zircaloy-2 cladding surface by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens then undergone heating process at 700oC, 900oC and 1100oC respectively in Argon gas atmosphere for 1 hour. The microstructure and crystal structure of the coated cladding were characterized by optical microscope and XRD respectively. The optical microscope showed the growth of the grains with increasing temperature. XRD examination on the specimens revealed that the ZrC crystal structure on the cladding surface occurred only at 1100oC, but it did not appear at 700oC and 900oC. It can be concluded that dipping process of specimen in colloidal graphite with subsequent heating at 1100oC provided ZrC film coated on zircaloy-2 cladding. The heating process at this temperature allowed carbon atoms to diffuse into zircaloy surface to form ZrC film. PELAPISAN ZrC PADA KELONGSONG ELEMEN BAKAR NUKLIR ZIRKALOI-2. Riset yang intensif pada elemen bakar reaktor berpendingin air dengan fraksi bakar tinggi terus dilakukan dalam rangka memperpanjang umur operasi elemen bakar. Salah satu riset tersebut berupa proses untuk mengubah kelongsong berbasis zirkonium yang ada saat ini dengan cara pelapisan. Cara ini diharapkan akan memperbaiki ketahanan korosi kelongsong tanpa mengubah dimensi kelongsong tersebut. Pada riset ini, lapisan tipis ZrC dilapiskan pada permukaan kelongsong zirkaloi-2 melalui proses pencelupan (dipping spesimen

  6. Vanadium diffusion coating on HT-9 cladding for mitigating the fuel cladding chemical interactions

    Energy Technology Data Exchange (ETDEWEB)

    Lo, Wei-Yang; Yang, Yong, E-mail: yongyang@ufl.edu

    2014-08-01

    Fuel cladding chemical interaction (FCCI) has been identified as one of the crucial issues for developing Ferritic/Martensitic (F/M) stainless steel claddings for metallic fuels in a fast reactor. The anticipated elevated temperature and high neutron flux can significantly aggravate the FCCI, in terms of formation of inter-diffusion and lower melting point eutectic phases. To mitigate the FCCI, vanadium carbide coating as a diffusion barrier was deposited on the HT-9 substrate using a pack cementation diffusion coating (PCDC) method, and the processing temperature was optimized down to 730 °C. A solid metallurgical bonding between the coating layer and substrate was achieved, and the coating is free from through depth cracks. The microstructural characterizations using SEM and TEM show a nanostructured grain structure. EDS/WDS and XRD analysis confirm the phase of coating layer as V{sub 2}C. Diffusion couple tests at 660 °C for 100 h demonstrate that V{sub 2}C layer with a thickness of less than 5 μm can effectively eliminate the inter-diffusion between the lanthanide cerium and HT-9 steel.

  7. Vanadium diffusion coating on HT-9 cladding for mitigating the fuel cladding chemical interactions

    Science.gov (United States)

    Lo, Wei-Yang; Yang, Yong

    2014-08-01

    Fuel cladding chemical interaction (FCCI) has been identified as one of the crucial issues for developing Ferritic/Martensitic (F/M) stainless steel claddings for metallic fuels in a fast reactor. The anticipated elevated temperature and high neutron flux can significantly aggravate the FCCI, in terms of formation of inter-diffusion and lower melting point eutectic phases. To mitigate the FCCI, vanadium carbide coating as a diffusion barrier was deposited on the HT-9 substrate using a pack cementation diffusion coating (PCDC) method, and the processing temperature was optimized down to 730 °C. A solid metallurgical bonding between the coating layer and substrate was achieved, and the coating is free from through depth cracks. The microstructural characterizations using SEM and TEM show a nanostructured grain structure. EDS/WDS and XRD analysis confirm the phase of coating layer as V2C. Diffusion couple tests at 660 °C for 100 h demonstrate that V2C layer with a thickness of less than 5 μm can effectively eliminate the inter-diffusion between the lanthanide cerium and HT-9 steel.

  8. Seismic behaviour of fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-11-01

    A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

  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. DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Kessinger, G.; Thompson, M.

    2009-08-07

    The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/L [U] and 1 M [H{sup +}] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 L of product solution, which was over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 M. A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of about 5000 {eta}Ci/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO{sub 2} present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U.

  11. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    Directory of Open Access Journals (Sweden)

    Shengli Chen

    2017-01-01

    Full Text Available Neutronic performance is investigated for a potential accident tolerant fuel (ATF, which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod, and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

  12. Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Sharon M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chattin, Marc Rhea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giaquinto, Joseph [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T2O. In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding

  13. Cr plating technology for preventing Fuel Cladding Chemical Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Ryu, Ho Jin; Jee, Seung Hyun; Cheon, Jin Sik; Lee, Byoung Oon; Lee, Chan Bock; Yang, Seong Woo [KAERI, Daejeon (Korea, Republic of)

    2010-11-15

    The objectives of the report are to analyze chrome electroplating technology in order to apply in the field of diffusion barrier to suppress Fuel-Cladding Chemical Interaction (FCCI). This report consists of the principle of the chrome electroplating, plating parameter and possibility of the barrier application. Chrome plating has been considered as one of the probable candidates in the field of barrier tube because of its simpleness, superior FCCI resistance, and effective coating performance at relatively low cost. However, cracks can be generate at the surface of the coating surface which reduces the coating performance. To minimize such a crack, controlling plating parameter like bath composition and bath temperature, current profile, and post-heat treatment has been reviewed. Concept for the application at the inner surface of the cladding has been also described. Based on the technology that suggested at the present report, optimizing plating parameter will be carried out. After the performance test like diffusion couple test of the metallic fuel, final barrier condition will be concluded and the fabrication of the prototype barrier tube will be conducted in the near future

  14. Screening of advanced cladding materials and UN-U3Si5 fuel

    Science.gov (United States)

    Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

    2015-07-01

    In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

  15. DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

    2002-04-01

    Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

  16. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1980-09-01

    Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U.

  17. Improvement in PCI property of PWR fuel cladding by texture control

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, S. (Kansai Electric Power Co., Inc., Osaka (Japan)); Abeta, S.; Ozawa, M.; Takahashi, T.

    1993-09-01

    Effects of texture on out-of-pile Stress Corrosion Cracking (SCC) resistance in Zircaloy fuel cladding tube and the Pellet-Clad Interaction (PCI) property of a fuel rod using texture controlled cladding tube under power ramp conditions are described. The cladding tube with radial texture, which means that the c-axis of hcp crystal of Zr is highly concentrated in the radial direction of the tube, showed excellent performance in out-of-pile SCC tests and power ramp tests. (author).

  18. Scratch Behaviors of Cr-Coated Zr-Based Fuel Claddings for Accident-Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Il-Hyun; Kim, Hyun-Gil; Kim, Hyung-Kyu; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As the progression of Fukushima accident is worsened by the runaway reaction at a high temperature above 1200 .deg. C, it is essential to ensure the stabilities of coating layers on conventional Zr-based alloys during normal operations as well as severe accident conditions. This is because the failures of coating layer result in galvanic corrosion phenomenon by potential difference between coating layer and Zr alloy. Also, it is possible to damage the coating layer during handling and manufacturing process by contacting structural components of a fuel assembly. So, adhesion strength is one of the key factors determining the reliability of the coating layer on conventional Zr-based alloy. In this study, two kinds of Cr-coated Zr-based claddings were prepared using arc ion plating (AIP) and direct laser (DL) coating methods. The objective is to evaluate the scratch deformation behaviors of each coating layers on Zr alloys. Large area spallation below normal load of about 15 N appeared to be the predominant mode of failure in the AIP coating during scratch test. However, no tensile crack were found in entire stroke length. In DL coating, small plastic deformation and grooving behavior are more dominant scratching results. It was observed that the change of the slope of the COF curve did not coincide with the failure of coating layer.

  19. Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

    Science.gov (United States)

    Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

    2016-10-01

    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

  20. Final data report for the instrumented fuel assembly (IFA)-432

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, E.R.; Cunningham, M.E.; Lanning, D.D.

    1982-06-01

    This report presents the in-reactor data collected during the irradiation of the six-rod instrumented fuel assembly (IFA)-432 in the Halden Boiling Water Reactor (HBWR) from June 1980 through June 1981. This Pacific Northwest Laboratory (PNL)-designed assembly was one of a series of US Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance computer codes. IFA-432 operated from December 1975 until June 1981, when it was removed from the reactor. Two of the rods were removed for examination, and the assembly was reinserted in December 1981 to obtain additional data. Fuel centerline temperatures, cladding elongations, internal fuel rod pressures, and local powers at thermocouple positions were monitored during the irradiation of IFA-432; and the resulting data are presented in this report.

  1. Neutronic evaluation of coating and cladding materials for accident tolerant fuels

    OpenAIRE

    Younker, I; Fratoni, M

    2016-01-01

    © 2015 Elsevier Ltd. All rights reserved. In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding inter...

  2. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  3. Compatibility study between U-UO2 cermet fuel and T91 cladding

    Science.gov (United States)

    Mishra, Sudhir; Kaity, Santu; Khan, K. B.; Sengupta, Pranesh; Dey, G. K.

    2016-12-01

    Cermet is a new fuel concept for the fast reactor system and is ideally designed to combine beneficial properties of both ceramic and metal. In order to understand fuel clad chemical compatibility, diffusion couples were prepared with U-UO2 cermet fuel and T91 cladding material. These diffusion couples were annealed at 923-1073 K for 1000 h and 1223 K for 50 h, subsequently their microstructures were examined using scanning electron microscope (SEM), X-ray energy dispersive spectroscope (EDS) and electron probe microanalyser (EPMA). It was observed that the interaction between the fuel and constituents of T91 clad was limited to a very small region up to the temperature 993 K and discrete U6(Fe,Cr) and U(Fe,Cr)2 intermetallic phases developed. Eutectic microstructure was observed in the reaction zone at 1223 K. The activation energy for reaction at the fuel clad interface was determined.

  4. Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, M.S.; Croucher, D.W.

    1979-01-01

    The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO/sub 2/-fuel during a PCM event.

  5. A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel

    Science.gov (United States)

    Lee, Young-Ho; Byun, Thak Sang

    2015-10-01

    Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate the tribological compatibilities of self-mated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

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

  7. Raman Spectroscopy Analysis of Oxide Film on Spent Fuel Rod Cladding from Qinshan PhaseⅠNPP

    Institute of Scientific and Technical Information of China (English)

    WANG; Hua-cai; TANG; Qi; FU; Cheng; LIANG; Zheng-qiang

    2015-01-01

    The outside surface of cladding is one of the important factors limiting the service life of the fuel rods.Studying the structure of oxide film under reactor operating conditions has great significance in study of the cause of different appearances of cladding,establishing the relationship between oxide film thickness and oxide structure

  8. U-Mo Foil/Cladding Interactions in Friction Stir Welded Monolithic RERTR Fuel Plates

    Energy Technology Data Exchange (ETDEWEB)

    D.D. Keiser; J.F. Jue; C.R. Clark

    2006-10-01

    Interaction between U-Mo fuel and Al has proven to dramatically impact the overall irradiation performance of RERTR dispersion fuels. It is of interest to better understand how similar interactions may affect the performance of monolithic fuel plates, where a uranium alloy fuel is sandwiched between aluminum alloy cladding. The monolithic fuel plate removes the fuel matrix entirely, which reduces the total surface area of the fuel that is available to react with the aluminum and moves the interface between the fuel and cladding to a colder region of the fuel plate. One of the major fabrication techniques for producing monolithic fuel plates is friction stir welding. This paper will discuss the interactions that can occur between the U-Mo foil and 6061 Al cladding when applying this fabrication technique. It has been determined that the time at high temperatures should be limited as much as is possible during fabrication or any post-fabrication treatment to reduce as much as possible the interactions between the foil and cladding. Without careful control of the fabrication process, significant interaction between the U-Mo foil and Al alloy cladding can result. The reaction layers produced from such interactions can exhibit notably different morphologies vis-à-vis those typically observed for dispersion fuels.

  9. Screening of advanced cladding materials and UN–U{sub 3}Si{sub 5} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R., E-mail: nbrown@bnl.gov; Todosow, Michael; Cuadra, Arantxa

    2015-07-15

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U{sub 3}Si{sub 5} fuels have the potential to exhibit reactor physics and fuel management performance similar to UO{sub 2}. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO{sub 2}) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO{sub 2} fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO{sub 2}–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding. The objective of the U{sub 3}Si{sub 5} phase in the UN–U{sub 3}Si{sub 5} fuel concept is to shield the nitride phase from water. It was shown that UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO{sub 2}–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to {sup 14}N content in UN ceramic composites is high

  10. 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

  11. 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)

    2014-12-30

    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

  12. Nuclear reactor composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  13. Characterization of Cassini GPHS Fueled-Clad Production Girth Welds

    Energy Technology Data Exchange (ETDEWEB)

    Franco-Ferreira, E.A.

    2000-03-23

    Fueled clads for radioisotope power systems are produced by encapsulating {sup 238}PuO{sub 2} in iridium alloy cups, which are joined at their equators by gas tungsten arc welding. Cracking problems at the girth weld tie-in area during production of the Galileo/Ulysses GPHS capsules led to the development of a first-generation ultrasonic test for girth weld inspection at the Savannah River Plant. A second-generation test and equipment with significantly improved sensitivity and accuracy were jointly developed by the Oak Ridge Y-12 Plant and Westinghouse Savannah River Company for use during the production of Cassini GPHS capsules by the Los Alamos National Laboratory. The test consisted of Lamb wave ultrasonic scanning of the entire girth weld from each end of the capsule combined with a time-of-flight evaluation to aid in characterizing nonrelevant indications. Tangential radiography was also used as a supplementary test for further evaluation of reflector geometry. Each of the 317 fueled GPHS capsules, which were girth welded for the Cassini Program, was subjected to a series of nondestructive tests that included visual, dimensional, helium leak rate, and ultrasonic testing. Thirty-three capsules were rejected prior to ultrasonic testing. Of the 44 capsules rejected by the standard ultrasonic test, 22 were upgraded to flight quality through supplementary testing for an overall process acceptance rate of 82.6%. No confirmed instances of weld cracking were found.

  14. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  15. Characterization of SiC-SiC composites for accident tolerant fuel cladding

    Science.gov (United States)

    Deck, C. P.; Jacobsen, G. M.; Sheeder, J.; Gutierrez, O.; Zhang, J.; Stone, J.; Khalifa, H. E.; Back, C. A.

    2015-11-01

    Silicon carbide (SiC) is being investigated for accident tolerant fuel cladding applications due to its high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. An engineered cladding design combining monolithic SiC and SiC-SiC composite layers could offer a tough, hermetic structure to provide improved performance and safety, with a failure rate comparable to current Zircaloy cladding. Modeling and design efforts require a thorough understanding of the properties and structure of SiC-based cladding. Furthermore, both fabrication and characterization of long, thin-walled SiC-SiC tubes to meet application requirements are challenging. In this work, mechanical and thermal properties of unirradiated, as-fabricated SiC-based cladding structures were measured, and permeability and dimensional control were assessed. In order to account for the tubular geometry of the cladding designs, development and modification of several characterization methods were required.

  16. Temperature limits for LMFBR fuel cladding under upset and emergency operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Govindarajan, S.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam Tamilnadu (India). Nuclear Systems Division

    1996-07-01

    LMFBR fuel pin cladding tube is subjected to high transient temperatures during incidents such as pump trip, pump to grid plate pipe rupture etc. It is required to know temperature limits under such transient operating conditions for components involved while analyzing such incidents. A methodology for deriving such limits for fuel clad tube is worked out in this paper by making use of the transient damage correlation proposed by W.F. Brizes et. al.

  17. Cold spray deposition of Ti2AlC coatings for improved nuclear fuel cladding

    Science.gov (United States)

    Maier, Benjamin R.; Garcia-Diaz, Brenda L.; Hauch, Benjamin; Olson, Luke C.; Sindelar, Robert L.; Sridharan, Kumar

    2015-11-01

    Coatings of Ti2AlC MAX phase compound have been successfully deposited on Zircaloy-4 (Zry-4) test flats, with the goal of enhancing the accident tolerance of LWR fuel cladding. Low temperature powder spray process, also known as cold spray, has been used to deposit coatings ∼90 μm in thickness using powder particles of accident tolerance to nuclear fuel cladding.

  18. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Allen, Todd; Cole, James

    2013-02-27

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

  20. Direct observation of fuel-cladding mechanical interaction (FCMI) in mixed-oxide fast reactor fuel pins

    Science.gov (United States)

    Foster, J. P.; Nayak, U. P.

    1981-10-01

    The WSA-1 and WSA-2 fuel pins exhibit experimental evidence of fuel-cladding mechanical interaction (FCMI) as a result of steady-state irradiation. The direct FCMI evidence involves a comparison of local axial and hoop mechanical strain profiles. The determination of the local axial mechanical strain was possible because of the placement of axial hardness marks 12.7 mm apart along a line parallel to the tubing axis spanning the fuel column. The measured cladding local axial and hoop mechanical deformations were the same within experimental error. The experimental results are in contrast to gas pressurized tube data which exhibit no axial mechanical deformation. A substantial amount of indirect evidence further illustrating the influence of FCMI on the cladding mechanical strain profile is also discussed. The conditions leading to steady-state FCMI are: high fuel smear density (i.e. low fuel-cladding gaps and/or high fuel pellet density), thin wall cladding, low cladding swelling and low fission gas pressure.

  1. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

  2. Simulation of accident and normal fuel rod work with Zr-cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tutnov, Anton A.; Tutnov, Alexander A. [Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.

    1995-12-31

    The technique of simulation of heat-physics, strength and safety characteristics of reactor RBMK and WWER rods under steady-state, transient and accident conditions is presented. That technique is used in mechanic and heat physics codes PULSAR-2 and STALACTITE. Simulation in both full scale and the most stress-loading part of cladding statement under accident conditions are considered. In this zone local swelling and cladding failure are possible. The accident simulation is based on the mechanical creep-plasticity problem solution in three-dimensional approach. The local cladding swelling is initiated with determining of little hot spot on the clad with several degrees temperature departure from average value. Mechanical problem is solved by finite elements method. Interaction of Zr with steam is taken in to account. Fuel and cladding melting, shortness and dispersion formation processes are simulated under subsequent rods warming up. (author). 2 refs., 6 figs.

  3. Initial Cladding Condition

    Energy Technology Data Exchange (ETDEWEB)

    E. Siegmann

    2000-08-22

    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&O 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

  4. Feasibility Study on the Sodium Compatibility Test for Fuel Cladding of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Shin, Sang Hun; Park, Sang Gyu; Ryu, Woo Seog; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A Sodium-cooled Fast Reactor (SFR), a reactor that uses fast neutrons as a fission process, is considered one of the most probable candidates in next-generation reactors because it can maximize the uranium utilization when compared to conventional water reactor. Liquid sodium is used as a coolant in a SFR, because it has superior efficiency of fast neutron economy and high thermal conductivity, which enables a high power core design. However, previous research reported that fuel cladding materials like austenitic and ferritic-martensitic steel (FMS) react sodium coolant so that it results in the loss of the thickness, intergranular attack, and carburization or decarburization process to induce the change of the mechanical property. Fuel cladding, a seamless tube which has approximately 0.5mm in thickness and 3m in length is the component which covers fuel to protect radioactive species from being released. Because of its smaller thickness, the mechanical properties of the cladding are easily affected by the small changes of material property. This paper summarizes the status of sodium-material compatibility facility and proposes the optimal option in the case of the SFR fuel cladding. Previous researches revealed that assessing in-situ mechanical property is important in the case of cladding material owing to its dimensional characteristic. Optimal test method for assessing sodium compatibility of the cladding tube can be proposed that pressurized creep test under the controlled liquid sodium environment.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V., E-mail: lalgudi@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Antunes, Renato A. [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Ramanathan, Lalgudi V. [Electrocell Ind. Com. Equip. Elet. LTDA (CIETEC), Sao Paulo, SP (Brazil)

    2013-07-01

    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)

  6. Radiation Effects Simulation of Fuel Assemblies

    Institute of Scientific and Technical Information of China (English)

    CUI; Yao

    2015-01-01

    Due to a large number of photons irradiated by the fuel assemblies after radiation in the reactor,the data acquisition and image reconstruction will be interfered seriously for the nuclear fuel assembly non-destructive testing system.Therefore,in process of the fuel assembly NDT system

  7. A micromechanical model for predicting hydride embrittlement in nuclear fuel cladding material

    Science.gov (United States)

    Chan, K. S.

    1996-01-01

    A major concern about nuclear fuel cladding under waste repository conditions is that the slow cooling rate anticipated in the repository may lead to the formation of excessive radial hydrides, and cause embrittlement of the cladding materials. In this paper, the development of a micromechanical model for predicting hydride-induced embrittlement in nuclear fuel cladding is presented. The important features of the proposed model are: (1) the capability to predict the orientation, morphology, and types of hydrides under the influence of key variables such as cooling rate, internal pressure, and time, and (2) the ability to predict the influence of hydride orientation and morphology on the tensile ductility and fracture toughness of the cladding material. Various model calculations are presented to illustrate the characteristics and utilities of the proposed methodology. A series of experiments was also performed to check assumptions used and to verify some of the model predictions.

  8. Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings.

  9. Early implementation of SiC cladding fuel performance models in BISON

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.

  10. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    OpenAIRE

    Bo Cheng; Young-Jin Kim; Peter Chou

    2016-01-01

    In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

  11. FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    2014-03-28

    The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.

  12. FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    2014-03-28

    The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.

  13. Development of a used fuel cladding damage model incorporating circumferential and radial hydride responses

    Science.gov (United States)

    Chen, Qiushi; Ostien, Jakob T.; Hansen, Glen

    2014-04-01

    At the completion of the fuel drying process, used fuel Zry4 cladding typically exhibits a significant population of δ-hydride inclusions. These inclusions are in the form of small platelets that are generally oriented both circumferentially and radially within the cladding material. There is concern that radially-oriented hydride inclusions may weaken the cladding material and lead to issues during used fuel storage and transportation processes. A high fidelity model of the mechanical behavior of hydrides has utility in both designing fuel cladding to be more resistant to this hydride-induced weakening and also in suggesting modifications to drying, storage, and transport operations to reduce the impact of hydride formation and/or the avoidance of loading scenarios that could overly stress the radial inclusions. We develop a mechanical model for the Zry4-hydride system that, given a particular morphology of hydride inclusions, allows the calculation of the response of the hydrided cladding under various loading scenarios. The model treats the Zry4 matrix material as J2 elastoplastic, and treats the hydrides as platelets oriented in predefined directions (e.g., circumferentially and radially). The model is hosted by the Albany analysis framework, where a finite element approximation of the weak form of the cladding boundary value problem is solved using a preconditioned Newton-Krylov approach. Instead of forming the required system Jacobian operator directly or approximating its action with a differencing operation, Albany leverages the Trilinos Sacado package to form the Jacobian via automatic differentiation. We present results that describe the performance of the model in comparison with as-fabricated Zry4 as well as HB Robinson fuel cladding. Further, we also present performance results that demonstrate the efficacy of the overall solution method employed to host the model.

  14. Development of a used fuel cladding damage model incorporating circumferential and radial hydride responses

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Qiushi, E-mail: qiushi@clemson.edu [Glenn Department of Civil Engineering, Clemson University, Clemson, SC 29634 (United States); Ostien, Jakob T., E-mail: jtostie@sandia.gov [Mechanics of Materials Dept. 8256, Sandia National Laboratories, P.O. Box 969, Livermore, CA 94551-0969 (United States); Hansen, Glen, E-mail: gahanse@sandia.gov [Computational Multiphysics Dept. 1443, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185-1321 (United States)

    2014-04-01

    At the completion of the fuel drying process, used fuel Zry4 cladding typically exhibits a significant population of δ-hydride inclusions. These inclusions are in the form of small platelets that are generally oriented both circumferentially and radially within the cladding material. There is concern that radially-oriented hydride inclusions may weaken the cladding material and lead to issues during used fuel storage and transportation processes. A high fidelity model of the mechanical behavior of hydrides has utility in both designing fuel cladding to be more resistant to this hydride-induced weakening and also in suggesting modifications to drying, storage, and transport operations to reduce the impact of hydride formation and/or the avoidance of loading scenarios that could overly stress the radial inclusions. We develop a mechanical model for the Zry4-hydride system that, given a particular morphology of hydride inclusions, allows the calculation of the response of the hydrided cladding under various loading scenarios. The model treats the Zry4 matrix material as J{sub 2} elastoplastic, and treats the hydrides as platelets oriented in predefined directions (e.g., circumferentially and radially). The model is hosted by the Albany analysis framework, where a finite element approximation of the weak form of the cladding boundary value problem is solved using a preconditioned Newton–Krylov approach. Instead of forming the required system Jacobian operator directly or approximating its action with a differencing operation, Albany leverages the Trilinos Sacado package to form the Jacobian via automatic differentiation. We present results that describe the performance of the model in comparison with as-fabricated Zry4 as well as HB Robinson fuel cladding. Further, we also present performance results that demonstrate the efficacy of the overall solution method employed to host the model.

  15. Stability increase of fuel clad with zirconium oxynitride thin film by metalorganic chemical vapor deposition

    Energy Technology Data Exchange (ETDEWEB)

    Jee, Seung Hyun [Department of Materials Science and Engineering, Yonsei University, 134 Sinchon Dong, Seoul 120-749 (Korea, Republic of); Materials Research and Education Center, Dept. of Mechanical Engineering, Auburn University, 275 Wilmore Labs, AL 36849-5341 (United States); Kim, Jun Hwan; Baek, Jong Hyuk [Recycled Fuel Development Division, Korea Atomic Energy Research Institute, P.O. Box 105, Yuseong, Daejeon, 305-600 (Korea, Republic of); Kim, Dong-Joo [Materials Research and Education Center, Dept. of Mechanical Engineering, Auburn University, 275 Wilmore Labs, AL 36849-5341 (United States); Kang, Seong Sik [Regulatory Research Division, Korea Institute of Nuclear Safety, 19, Guseong-Dong, Yuseong-Gu, Daejeon, 305-338 (Korea, Republic of); Yoon, Young Soo, E-mail: yoonys@yonsei.ac.kr [Department of Materials Science and Engineering, Yonsei University, 134 Sinchon Dong, Seoul 120-749 (Korea, Republic of)

    2012-06-01

    A zirconium oxynitride (ZON) thin film was deposited onto HT9 steel as a cladding material by a metalorganic chemical vapor deposition (MOCVD) in order to prevent a fuel-clad chemical interaction (FCCI) between a U-10 wt% Zr metal fuel and a clad material. X-ray diffraction spectrums indicated that the mixture of structures of zirconium nitride, oxide and carbide in the MOCVD grown ZON thin films. Also, typical equiaxial grain structures were found in plane and cross sectional images of the as-deposited ZON thin films with a thickness range of 250-500 nm. A depth profile using auger electron microscopy revealed that carbon and oxygen atoms were decreased in the ZON thin film deposited with hydrogen gas flow. Diffusion couple tests at 800 Degree-Sign C for 25 hours showed that the as-deposited ZON thin films had low carbon and oxygen content, confirmed by the Energy Dispersive X-ray Spectroscopy, which showed a barrier behavior for FCCI between the metal fuel and the clad. This result suggested that ZON thin film cladding by MOCVD, even with the thickness below the micro-meter level, has a high possibility as an effective FCCI barrier. - Highlights: Black-Right-Pointing-Pointer Zirconium oxynitride (ZON) deposited by metal organic chemical vapor deposition. Black-Right-Pointing-Pointer Prevention of fuel cladding chemical interaction (FCCI) investigated. Black-Right-Pointing-Pointer Interfusion reduced by between metal fuel (U-10 wt% Zr) and a HT9 cladding material. Black-Right-Pointing-Pointer Hydrogenation of the ZON during growth improved the FCCI barrier performance.

  16. Demonstration of fuel resistant to pellet-cladding interaction. Phase 2. First semiannual report, January-June 1979. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1979-08-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress and reactive fission products during reactor service. This is the first semiannual progress report for Phase 2 of this program (January-June 1979). Progress in the irradiation testing of barrier fuel and of unfueled barrier cladding specimens is reported.

  17. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

  18. Cladding stress during extended storage of high burnup spent nuclear fuel

    Science.gov (United States)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  19. Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J.C.

    1985-08-01

    Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.

  20. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

    Directory of Open Access Journals (Sweden)

    Daxin Gong

    2015-01-01

    Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

  1. Two dimensional structural analysis of reactor fuel element claddings due to local effects

    Energy Technology Data Exchange (ETDEWEB)

    Karimi, R; Wolf, L

    1978-04-01

    Two dimensional thermoelastic and inelastic stresses and deformation of typical LWR (PWR) and LMFBR (CRBR) claddings are evaluated by utilizing the following codes, for (1) Thermoelastic analysis (a) STRESS Code (b) SEGPIPE Code (2) Thermoinelastic analysis (a) Modified version of the GOGO code (b) One dimensional GRO-II code. The primary objective of this study is to analyze the effect of various local perturbations in the clad temperature field, namely eccentrically mounted fuel pellet, clad ovality, power tilt across the fuel and clad-coolant heat transfer variation on the cladding stress and deformation. In view of the fact that the thermoelastic analysis is always the first logical choice entering the structural field, it was decided to start the analysis with the two dimensional codes such as STRESS and SEGPIPE. Later, in order to assess the validity and compare the thermoelastic results to those obtained for actual reactor conditions, a two dimensional code, namely a modified version of the GOGO code, was used to account for inelastic effects such as irradiation and thermal creep and swelling in the evaluation. The comparison of thermoelastic and inelastic results shows that the former can be used effectively to analyze LWR fuel pin over 350 hours of lifetime under the most adverse condition and 500 hours of lifetime for an LMFBR fuel pin. Beyond that the inelastic solution must be used. The impact of the individual thermal perturbation and combinations thereof upon the structural quantity is also shown. Finally, the effect of rod displacement on the two dimensional thermal and structural quantities of the LMFBR fuel pin cladding is analyzed.

  2. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jason D. Hales; Various

    2014-09-01

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  3. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Courty, Olivier, E-mail: o.courty@gmail.com [Pennsylvania State University, 45 Bd Gouvion Saint Cyr, 75017 Paris (France); Motta, Arthur T., E-mail: atm2@psu.edu [Department of Mechanical and Nuclear Engineering, 227 Reber Building, Penn State University, University Park, PA 16802 (United States); Hales, Jason D., E-mail: jason.hales@inl.gov [Fuels Modeling and Simulation Department, Idaho National Laboratory (United States)

    2014-09-15

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  4. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    Science.gov (United States)

    Courty, Olivier; Motta, Arthur T.; Hales, Jason D.

    2014-09-01

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick's law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  5. Fuel Cell Electrodes for Hydrogen-Air Fuel Cell Assemblies.

    Science.gov (United States)

    The report describes the design and evaluation of a hydrogen-air fuel cell module for use in a portable hydrid fuel cell -battery system. The fuel ... cell module consists of a stack of 20 single assemblies. Each assembly contains 2 electrically independent cells with a common electrolyte compartment

  6. Optimization of N18 Zirconium Alloy for Fuel Cladding of Water Reactors

    Institute of Scientific and Technical Information of China (English)

    B.X. Zhou; M. Y. Yao; Z.K. Li; X.M. Wang; J. Zhoua; C.S. Long; Q. Liu; B.F. Luan

    2012-01-01

    In order to optimize the microstructure and composition of N18 zirconium alloy (Zr-1Sn-0.35Nb-0.35Fe-0.1Cr, in mass fraction, %), which was developed in China in 1990s, the effect of microstructure and composition variation on the corrosion resistance of the N18 alloy has been investigated. The autoclave corrosion tests were carried out in super heated steam at 400 ~C/10.3 MPa, in deionized water or lithiated water with 0.01 mol/L LiOH at 360 ~C/18.6 MPa. The exposure time lasted for 300-550 days according to the test temperature. The results show that the microstructure with a fine and uniform distribution of second phase particles (SPPs), and the decrease of Sn content from 1% (in mass fraction, the same as follows) to 0.8% are of benefit to improving the corrosion resistance; It is detrimental to the corrosion resistance if no Cr addition. The addition of Nb content with upper limit (0.35%) is beneficial to improving the corrosion resistance. The addition of Cu less than 0.1% shows no remarkable influence upon the corrosion resistance for N18 alloy. Comparing the corrosion resistance of the optimized N18 with other commercial zirconium alloys, such as Zircaloy-4, ZIRLO, E635 and Ell0, the former shows superior corrosion resistance in all autoclave testing conditions mentioned above. Although the data of the corrosion resistance as fuel cladding for high burn-up has not been obtained yet, it is believed that the optimized N18 alloy is promising for the candidate of fuel cladding materials as high burn-up fuel assemblies. Based on the theory that the microstructural evolution of oxide layer during corrosion process will affect the corrosion resistance of zirconium alloys, the improvement of corrosion resistance of the N18 alloy by obtaining the microstructure with nano-size and uniform distribution of SPPs, and by decreasing the content of Sn and maintaining the content of Cr is discussed.

  7. 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

  8. BISON Investigation of the Effect of the Fuel- Cladding Contact Irregularities on the Peak Cladding Temperature and FCCI Observed in AFC-3A Rodlet 4

    Energy Technology Data Exchange (ETDEWEB)

    Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    The primary objective of this report is to document results of BISON analyses supporting Fuel Cycle Research and Development (FCRD) activities. Specifically, the present report seeks to provide explanation for the microstructural features observed during post irradiation examination of the helium-bonded annular U-10Zr fuel irradiated during the AFC-3A experiment. Post irradiation examination of the AFC-3A rodlet revealed microstructural features indicative of the fuel-cladding chemical interaction (FCCI) at the fuel-cladding interface. Presence of large voids was also observed in the same locations. BISON analyses were performed to examine stress and temperature profiles and to investigate possible correlation between the voids and FCCI. It was found that presence of the large voids lead to a formation of circumferential temperature gradients in the fuel that may have redirected migrating lanthanides to the locations where fuel and cladding are in contact. Resulting localized increase of lanthanide concentration is expected to accelerate FCCI. The results of this work provide important guidance to the post irradiation examination studies. Specifically, the hypothesis of lanthanides being redirected from the voids to the locations where the fuel and the cladding are in contact should be verified by conducting quantitative electron microscopy or Electron Probe Micro-Analyzer (EPMA). The results also highlight the need for computer models capable of simulating lanthanide diffusion in metallic fuel and establish a basis for validation of such models.

  9. Construction of in-situ creep strain test facility for the SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel (12CrMoVW) is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92(12Cr-2W) are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.

  10. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  11. 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)

  12. LMFBR fuel assembly design for HCDA fuel dispersal

    Science.gov (United States)

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  13. Surface modification techniques for increased corrosion tolerance of zirconium fuel cladding

    Science.gov (United States)

    Carr, James Patrick, IV

    Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively employed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neutron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objectives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and considering materials and processes for modifying the surface of zircaloy fuel cladding. This innovative research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their microstructural and

  14. High temperature nanoindentation hardness and Young's modulus measurement in a neutron-irradiated fuel cladding material

    Science.gov (United States)

    Kese, K.; Olsson, P. A. T.; Alvarez Holston, A.-M.; Broitman, E.

    2017-04-01

    Nanoindentation, in combination with scanning probe microscopy, has been used to measure the hardness and Young's modulus in the hydride and matrix of a high burn-up neutron-irradiated Zircaloy-2 cladding material in the temperature range 25-300 °C. The matrix hardness was found to decrease only slightly with increasing temperature while the hydride hardness was essentially constant within the temperature range. Young's modulus decreased with increasing temperature for both the hydride and the matrix of the high burn-up fuel cladding material. The hydride Young's modulus and hardness were higher than those of the matrix in the temperature range.

  15. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  16. The state of the art report on the development of advanced nuclear fuel cladding tube

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

    Since the operating conditions of modern PWR trend toward long-term operation, high burn-up, high coolant temperature and high pH, the need to develop a new advanced nuclear fuel cladding as an alternative to Zircaloy-4 increased. To overcome this problem, a number of researches to develop a advanced nuclear fuel cladding tube with superior corrosion resistance and creep resistance have been performed in many advanced nations in the field of nuclear power. Especially, some advanced cladding tubes are already confirmed to have an excellent in-pile properties from the test results in commercial reactor. Also in Korea, KAERI has been researching extensively to develop a high burn-up nuclear fuel cladding Zr alloy since 1990. To design new alloys, it is necessary to study the state of the art on the development of advanced alloys in other countries. In this report, as a part of development of advanced Zr alloy, we studied the state of the art on the development of ZIRLO in U.S.A., E635 in Russia, M5 in France, and MDA and NDA in Japan, which will be applied as basic data to develop an advanced Zr alloy.

  17. 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.

  18. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  19. Modelling anelastic contribution to nuclear fuel cladding creep and stress relaxation

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville, E-mail: ville.tulkki@vtt.fi; Ikonen, Timo

    2015-10-15

    In fuel behaviour modelling accurate description of the cladding mechanical response is important for both operational and safety considerations. While accuracy is desired, a certain level of simplicity is needed as both computational resources and detailed information on properties of particular cladding may be limited. Most models currently used in the integral codes divide the mechanical response into elastic and viscoplastic contributions. These have difficulties in describing both creep and stress relaxation, and often separate models for the two phenomena are used. In this paper we implement anelastic contribution to the cladding mechanical model, thus enabling consistent modelling of both creep and stress relaxation. We show that the model based on assumption of viscoelastic behaviour can be used to explain several experimental observations in transient situations and compare the model to published set of creep and stress relaxation experiments performed on similar samples. Based on the analysis presented we argue that the inclusion of anelastic contribution to the cladding mechanical models provides a way to improve the simulation of cladding behaviour during operational transients.

  20. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    Directory of Open Access Journals (Sweden)

    Bo Cheng

    2016-02-01

    Full Text Available In severe loss of coolant accidents (LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are

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

    Energy Technology Data Exchange (ETDEWEB)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Suga, Masataka [Kokan Keisoku Co., Kawasaki, Kanagawa (Japan)

    2001-03-01

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

  2. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Loss of Coolant Accident LWR Light Water Reactor MOX Mixed Oxide Fuel MTC Moderator Temperature Coefficient MWd/kgIHM Megawatt days per...working only with UO2 and UO2/PuO2 mixed oxide ( MOX ) fuels. 3.1 Studsvik Core Management Software CASMO-4E and SIMULATE-3 are the primary computational

  3. 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.

  4. A deformation and thermodynamic model for hydride precipitation kinetics in spent fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.

    1989-10-01

    Hydrogen is contained in the Zircaloy cladding of spent fuel rods from nuclear reactors. All the spent fuel rods placed in a nuclear waste repository will have a temperature history that decreases toward ambient; and as a result, most all of the hydrogen in the Zircaloy will eventually precipitate as zirconium hydride platelets. A model for the density of hydride platelets is a necessary sub-part for predicting Zircaloy cladding failure rate in a nuclear waste repository. A model is developed to describe statistically the hydride platelet density, and the density function includes the orientation as a physical attribute. The model applies concepts from statistical mechanics to derive probable deformation and thermodynamic functionals for cladding material response that depend explicitly on the hydride platelet density function. From this model, hydride precipitation kinetics depend on a thermodynamic potential for hydride density change and on the inner product of a stress tensor and a tensor measure for the incremental volume change due to hydride platelets. The development of a failure response model for Zircaloy cladding exposed to the expected conditions in a nuclear waste repository is supported by the US DOE Yucca Mountain Project. 19 refs., 3 figs.

  5. Measurement of Nucleate Pool Boiling Heat Transfer Limit using Fuel Cladding Material

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young; Shin, Chang Hwan; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    Zircaloy has been widely used as a fuel cladding material of light water reactor for more than three decades because it has a lower neutron absorption cross section and cracking rate. Recently, HANA-6 has been developed in KAERI (Korea Atomic Energy Research Institute) as the advanced fuel cladding for high burn-up fuel. Generally, under the normal and accident operating conditions of a nuclear reactor, the nuclear fuel cladding of zirconium based alloys undergoes the surface change, and the oxide layer can be formed. In such a case, the previous CHF correlations should be assessed and examined using the experimental results for not a fresh zircaloy surface but an oxidized one, to predict and examine the thermal margin and safety of a nuclear reactor core. Therefore, the experimental data using the oxidized zircaloy surface need to be provided quantitatively. In this paper, the CHF in saturated water pool boiling is measured and discussed using the specimens of zircaloy-4, HANA-6, and oxidized zircaloy-4 in high temperature air environment. The CHF of zircaloy-4, HANA-6, and oxidized surface was tested. Zircaloy-4 and HANA-6 had a similar CHF performance. This is because both are the zirconium based alloys, and appear the almost same water contact angle. On the other hands, the oxidized specimen became to be higher CHF than plain zircaloy-4 and HANA-6 specimens, due to smaller water contact angle (i. e., good hydrophilicity of specimen). The Kandlikar's (2001) correlation reasonably predicted the present experimental data.

  6. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

    2015-08-15

    Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

  7. Rod internal pressure of spent nuclear fuel and its effects on cladding degradation during dry storage

    Science.gov (United States)

    Kim, Ju-Seong; Hong, Jong-Dae; Yang, Yong-Sik; Kook, Dong-Hak

    2017-08-01

    Temperature and hoop stress limits have been used to prevent the gross rupture of spent nuclear fuel during dry storage. The stress due to rod internal pressure can induce cladding degradation such as creep, hydride reorientation, and delayed hydride cracking. Creep is a self-limiting phenomenon in a dry storage system; in contrast, hydride reorientation and delayed hydride cracking are potential degradation mechanisms activated at low temperatures when the cladding material is brittle. In this work, a conservative rod internal pressure and corresponding hoop stress were calculated using FRAPCON-4.0 fuel performance code. Based on the hoop stresses during storage, a study on the onset of hydride reorientation and delayed hydride cracking in spent nuclear fuel was conducted under the current storage guidelines. Hydride reorientation is hard to occur in most of the low burn-up fuel while some high burn-up fuel can experience hydride reorientation, but their effect may not be significant. On the other hand, delayed hydride cracking will not occur in spent nuclear fuel from pressurized water reactor; however, there is a lack of confirmatory data on threshold intensity factor for delayed hydride cracking and crack size distribution in the fuel.

  8. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  9. Irradiated MTR fuel assemblies sipping test

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, J.A.; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1997-10-01

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab.

  10. PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    Hales, J. D.; Gamble, K. A.

    2015-09-01

    Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.

  11. Evaluation of corrosion on the fuel performance of stainless steel cladding

    OpenAIRE

    de Souza Gomes Daniel; Abe Alfredo; Silva Antonio Teixeira e; Giovedi Claudia; Martins Marcelo Ramos

    2016-01-01

    In nuclear reactors, the use of stainless steel (SS) as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to a...

  12. Potential corrosion and degradation mechanisms of Zircaloy cladding on spent nuclear fuel in a tuff repository

    Energy Technology Data Exchange (ETDEWEB)

    Rothman, A.J.

    1984-09-01

    A literature review and analysis were made of corrosion and degradation processes applicable to Zircaloy cladding on spent nuclear fuel in a tuff repository. In particular, lifetime sought for the Zircaloy is 10,000 years. Among the potential failure mechanisms examined were: oxidation by steam, air, and water, including the effects of ions whose presence is anticipated in the water; mechanical overload; stress (creep) rupture; stress-corrosion cracking (SCC); and delayed failure due to hydride cracking. The conclusion is that failure due to oxidation is not credible, although a few experiments are suggested to confirm the effect of aqueous fluoride on the Zircaloy cladding. Mechanical overload is not a problem, and failure from stress-rupture does not appear likely based on a modified Larson-Miller analysis. Analysis shows that delayed hydride cracking is not anticipated for the bulk of spent fuel pins. However, for a minority of pins under high stress, there is some uncertainty in the analysis as a result of: (1) uncertainty about crack depths in spent fuel claddings and (2) the effect of slow cooling on the formation of radially oriented hydride precipitates. Experimental resolution is called for. Finally, insufficient information is currently available on stress-corrosion cracking. While evidence is presented that SCC failure is not likely to occur, it is difficult to demonstrate this conclusively because the process is not clearly understood and data are limited. Further experimental work on SCC susceptibility is especially needed.

  13. Synthesis of the Novel MAX Phases for the Future Nuclear Fuel Cladding and Structural Materials

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hyeok [Kyunghee Univ., Yongin (Korea, Republic of); Kim, Taehee; Lee, Taegyu; Ryu, H. J. [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    With these properties, the MAX phases are expected to be used for the Accident Tolerant Fuel (ATF) cladding and oxidation/corrosion resistance materials. Especially, the MAX phase can be used for the Gen-IV, SFR and HTGR, component materials which have to possess the thermal and corrosion resistance. The zirconium has been used to the nuclear industry for fuel cladding because of the small thermal neutron cross-section. Zr-based MAX phase was discovered by group Lapauw et al. They observed the Zr{sub 2}AlC and Zr{sub 3}AlC{sub 2} with the X-ray diffraction (XRD) patterns and backscattered electron detector. Fabrication of the Zr-containing MAX phase was investigated for nuclear fuel cladding and structural materials applications. A MAX phase with the Zr{sub 3}AlC{sub 2} structure was synthesized by spark plasma sintering of a powder mixture targeting (Zr{sub 0.5}Cr{sub 0.5}){sub 4}AlC{sub 3}. The formation of MAX phases was confirmed by XRD and EDS of sintered samples. In the future work, the electron probe micro analyzer (EPMA) and transmission electron microscopy (TEM) are required to certain analyze the elements composition and formation of the MAX phase.

  14. Mechanical behavior of aluminum-bearing ferritic alloys for accident-tolerant fuel cladding applications

    Science.gov (United States)

    Guria, Ankan

    Nuclear power currently provides about 13% of electrical power worldwide. Nuclear reactors generating this power traditionally use Zirconium (Zr) based alloys as the fuel cladding material. Exothermic reaction of Zr with steam under accident conditions may lead to production of hydrogen with the possibility of catastrophic consequences. Following the Fukushima-Daiichi incident, the exploration of accident-tolerant fuel cladding materials accelerated. Aluminum-rich (around 5 wt. %) ferritic steels such as Fecralloy, APMT(TM) and APM(TM) are considered as potential materials for accident-tolerant fuel cladding applications. These materials create an aluminum-based oxide scale protecting the alloy at elevated temperatures. Tensile deformation behavior of the above alloys was studied at different temperatures (25-500 °C) at a strain rate of 10-3 s-1 and correlated with microstructural characteristics. Higher strength and decent ductility of APMT(TM) led to further investigation of the alloy at various combination of strain rates and temperatures followed by fractography and detailed microscopic analyses. Serrations appeared in the stress-strain curves of APMT(TM) and Fecralloy steel tested in a limited temperature range (250-400 °C). The appearance of serrations is explained on the basis of dynamic strain aging (DSA) effect due to solute-dislocation interactions. The research in this study is being performed using the funds received from the US DOE Office of Nuclear Energy's Nuclear Energy University Programs (NEUP).

  15. Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement.

    Science.gov (United States)

    Škarohlíd, Jan; Ashcheulov, Petr; Škoda, Radek; Taylor, Andrew; Čtvrtlík, Radim; Tomáštík, Jan; Fendrych, František; Kopeček, Jaromír; Cháb, Vladimír; Cichoň, Stanislav; Sajdl, Petr; Macák, Jan; Xu, Peng; Partezana, Jonna M; Lorinčík, Jan; Prehradná, Jana; Steinbrück, Martin; Kratochvílová, Irena

    2017-07-25

    In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100-170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

  16. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    Science.gov (United States)

    Marcum, W. R.; Wachs, D. M.; Robinson, A. B.; Lillo, M. A.

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers.

  17. Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Sharon M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chattin, Marc Rhea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giaquinto, Joseph M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

  18. Establishment of China Nuclear Fuel Assembly Database

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; LIUTing-jin; JINYong-li

    2003-01-01

    During researching, designing, manufacturing and post irradiation, a large amount of data on fuel assembly of China nuclear power plants has been accumulated. It is necessary to collect the data together,so that the researchers, designers, manufactures and managers could use the data conveniently. It was proposed to establish a China Nuclear Fuel Assembly Database through the Internet on workstations during the year of 2003 to 2006, so the data would be shared in China nuclear industry.

  19. Conceptual design of ASTRID fuel sub-assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Beck, Thierry, E-mail: thierry.beck@cea.fr [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Blanc, Victor; Escleine, Jean-Michel [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Haubensack, David [CEA Cadarache, DEN, DER, F-13108 Saint-Paul-lez-Durance (France); Pelletier, Michel; Phelip, Mayeul [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Perrin, Benoît [AREVA-NP, 10 rue J. Récamier, 69456 Lyon Cedex 06 (France); Venard, Christophe [CEA Cadarache, DEN, DER, F-13108 Saint-Paul-lez-Durance (France)

    2017-04-15

    Highlights: • The fuel sub-assembly design for the ASTRID CFV core is described. • Innovative design choices have been made to comply with the GEN IV objectives. • The heterogeneous and the large fuel pins contribute to a low sodium void worth. • The upper neutron shielding is removable from the S/A head before washing. - Abstract: The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project has reached the end of its Conceptual Design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phénix and SuperPhénix Sodium Fast Reactors. The main objective to improve safety compared with current GEN II or III reactors led to a core design that demonstrates intrinsically safe behaviour. A negative sodium void worth is achieved thanks to a new fuel sub-assembly design including (U,Pu)O{sub 2} and UO{sub 2} axially heterogeneous fuel pins, a large cladding/small spacer wire bundle, a sodium plenum above the fuel pins, and upper neutron shielding with both enriched and natural boron carbide (B{sub 4}C) which also maintain a low secondary sodium activity level. As these Na-bonded B{sub 4}C pins can lead to the retention of unacceptable amounts of sodium, the whole upper neutron shielding has been made removable on-line through the sub-assembly head just before the washing operations. Finite elements calculations have been performed to increase the stiffness of the stamped spacer pads in order to analyse its effect on the core mechanical behaviour during hypothetical radial core flowering and compaction events. More generally, all design choices for ASTRID have been made with the permanent objective of minimising the sub-assembly height to decrease the overall costs of the reactor and the fuel cycle. This paper describes the fuel sub-assembly design for

  20. Evaluation of corrosion on the fuel performance of stainless steel cladding

    Directory of Open Access Journals (Sweden)

    de Souza Gomes Daniel

    2016-01-01

    Full Text Available In nuclear reactors, the use of stainless steel (SS as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4 under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion.

  1. Hydrogen uptake in Zircaloy-2 reactor fuel claddings studied with elastic recoil detection

    Science.gov (United States)

    Rajasekhara, S.; Doyle, B. L.; Enos, D. G.; Clark, B. G.

    2013-04-01

    The recent trend towards a high burn-up discharge spent nuclear fuel necessitates a thorough understanding of hydrogen uptake in Zr-based cladding materials that encapsulate spent nuclear fuel. Although it is challenging to experimentally replicate exact conditions in a nuclear reactor that lead to hydrogen uptake in claddings, in this study we have attempted to understand the kinetics of hydrogen uptake by first electrolytically charging Zircaloy-2 (Zr-2) cladding material for various durations (100 to 2,600 s), and subsequently examining hydrogen ingress with elastic recoil detection (ERD) and transmission electron microscopy (TEM). To understand the influence of irradiation damage defects on hydrogen uptake, an analogous study was performed on ion - irradiated (0.1, 1 and 25 dpa) Zr-2. Analysis of ERD data from the un-irradiated Zr-2 suggests that the growth of the hydride layer is diffusion controlled, and preliminary TEM results support this assertion. In un-irradiated Zr-2, the diffusivity of hydrogen in the hydride phase was found to be approximately 1.1 × 10-11 cm2/s, while the diffusivity in the hydride phase for lightly irradiated (0.1 and 1 dpa) Zr-2 is an order of magnitude lower. Irradiation to 25 dpa results in a hydrogen diffusivity that is comparable to the un-irradiated Zr-2. These results are compared with existing literature on hydrogen transport in Zr - based materials.

  2. Obtention of fracture properties of unirradiated fuel cladding from ring compression tests

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos Profesor Aranguren s/n, E-28040 Madrid (Spain); Consejo de Seguridad Nuclear (CSN), Justo Dorado 11, E-28040 Madrid (Spain); Gomez, F.J.; Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2009-06-15

    Zirconium alloy cladding is used as the first structural barrier to contain the nuclear fuel and the fission products. In addition to its neutron transparency, this material has a good corrosion resistance and remarkable mechanical properties at operational temperatures. Consequently, it is or paramount importance to precisely characterize the mechanical behaviour and fracture properties of irradiated cladding to ensure a safe operation. It is known that the mechanical behaviour of unirradiated zirconium alloy cladding is anisotropic. The elastoplastic response depends on the direction, namely radial, hoop or longitudinal. For this reason, different fracture properties should be expected in each direction. From the various tests employed to characterize the mechanical behaviour along the hoop direction in nuclear fuel cladding, the ring compression test is particularly useful to study material fracture. With this test it is possible to determine the moment when a real crack is formed, due to a sudden decrease in the applied load at a given displacement value. The aim of this research is to determine as precisely as possible the value of the fracture energy from the ring compression test load vs. displacement curves. To this end, a finite element calculation incorporating the cohesive zone model was performed. In this case, the cohesive zone theory is applied in its simplest form. It is considered that the cohesive crack transfers a constant stress until the displacement of this cohesive crack reaches a critical value. At this precise moment a real crack is generated. The properties of the softening curve of the cohesive zone model can be obtained by directly comparing the experimental load vs. displacement records with the finite element calculations. The area under the softening curve is the fracture energy, which is directly related with the material fracture toughness. The experimental data used in this work have been obtained on unirradiated Zirlo cladding

  3. Composite nozzle design for reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Marlatt, G.R.; Allison, D.K.

    1984-01-24

    A composite nozzle is described for a fuel assembly adapted for installation on the upper or lower end thereof and which is constructed from two components. The first component includes a casting weldment or forging designed to carry handling loads, support fuel assembly weight and flow loads, and interface with structural members of both the fuel assembly and reactor internal structures. The second component of the nozzle consists of a thin stamped bore machine flow plate adapted for attachment to the casting body. The plate is designed to prevent fuel rods from being ejected from the core and provide orifices for coolant flow to a predetermined value and pressure drop which is consistent with the flow at other locations in the core.

  4. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    OpenAIRE

    Shengli Chen; Cenxi Yuan

    2017-01-01

    Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into con...

  5. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    Science.gov (United States)

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  6. Polymer electrolyte membrane assembly for fuel cells

    Science.gov (United States)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2002-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  7. Nuclear reactor composite fuel assembly. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, D.M.; Cappiello, M.W.; Marr, D.R.; Omberg, R.P.

    1980-11-25

    A core and composite fuel assembly are described for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  8. 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

  9. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  10. GEH-4-63, 64: Proposal for irradiation of production brazed Zircaloy-2 clad fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, J.C.

    1961-05-18

    A brazed end closure is currently being used on prototypical NPR fuel elements. The production closure will use a braze alloy composed of 5% Be + 95% Zry-2 to braze the Zircaloy-2 cap to the jacket and to the metallic uranium core. A similar MTR test, a GEH-4-57, 58, used a braze alloy of the composition 4% Be + 12% Fe + 84% Zry-2 which melts at a lower temperature. In this previous test, element GEH-4-57 failed through a cladding defect located at the base of the braze heat affected zone. Because of this failure it would be desirable to subject a fuel element, which had been subjected to more severe brazing conditions, to the same conditions as GEH-4-57, 58. For this reason the thermal conditions of this test essentially match those of GEH-4-57, 58. This irradiation test consists of two identical fuel elements. The fuel material is normal metallic uranium, Zircaloy-2 clad of the tubular geometry, NPR inner size. The fuel was coextruded at Hanford by General Electric`s Fuels Preparation Department. Each element is 10.8 inches in length with flat Zircaloy-2 end caps brazed to the jacket and uranium core with the 5 Be + 95 Zry-2 brazing alloy, then TIG welded to further insure closure integrity. The elements ar 1.254 inches OD and 0.439 inches ID. For hydraulic purposes a 0.343 inch diamater flow restrictor has been fitted into the central flow channel of both elements.

  11. U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    George W. Griffith

    2011-10-01

    A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows for ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.

  12. RF Plasma Torch System for Metal Matrix Composite Production in Nuclear Fuel Cladding

    Science.gov (United States)

    Holik, Eddie, III

    2007-10-01

    For the first time in 30 years, plans are afoot to build new fission power plants in the US. It is timely to develop technology that could improve the safety and efficiency of new reactors. A program of development for advanced fuel cycles and Generation IV reactors is underway. The path to greater efficiency is to increase the core operating temperature. That places particular challenges to the cladding tubes that contain the fission fuel. A promising material for this purpose is a metal matrix composite (MMC) in which ceramic fibers are bonded within a high-strength steel matrix, much like fiberglass. Current MMC technology lacks the ability to effectively bond traditional high-temperature alloys to ceramic strands. The purpose of this project is to design an rf plasma torch system to use titanium as a buffer between the ceramic fibers and the refractory outer material. The design and methods of using an rf plasma torch to produce a non-equilibrium phase reaction to bond together the MMC will be discussed. The effects of having a long lived fuel cladding in the design of future reactors will also be discussed.

  13. Estimation of ring tensile properties of steam oxidized Zircaloy-4 fuel cladding under simulated LOCA condition

    Science.gov (United States)

    Shriwastaw, R. S.; Sawarn, Tapan K.; Banerjee, Suparna; Rath, B. N.; Dubey, J. S.; Kumar, Sunil; Singh, J. L.; Bhasin, Vivek

    2017-09-01

    The present study involves the estimation of ring tensile properties of Indian Pressurised Heavy Water Reactor (IPHWR) fuel cladding made of Zircaloy-4, subjected to experiments under a simulated loss-of-coolant-accident (LOCA) condition. Isothermal steam oxidation experiments were conducted on clad tube specimens at temperatures ranging from 900 to 1200 °C at an interval of 50 °C for different soaking periods with subsequent quenching in water at ambient temperature. The specimens, which survived quenching, were then subjected to ambient temperature ring tension test (RTT). The microstructure was correlated with the mechanical properties. The yield strength (YS) and ultimate tensile strength (UTS) increased initially with rise in oxidation temperature and time duration but then decreased with further increase in oxidation. Ductility is adversely affected with rising oxidation temperature and longer holding time. A higher fraction of load bearing phase and lower oxygen content in it ensures higher residual ductility. Cladding shows almost zero ductility behavior in RIT when load bearing phase fraction is less than 0.72 and its average oxygen concentration is greater than 0.58 wt%.

  14. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  15. New method to calculate the mechanical properties of unirradiated fuel cladding from ring tensile tests

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain); Consejo de Seguridad Nuclear (CSN), Justo Dorado 11, E-28040 Madrid (Spain); Gomez, F.J.; Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2009-06-15

    Nuclear fuel cladding is the first barrier used to confine the fuel and the fission products produced during irradiation. Zirconium alloys are used for this purpose due to their remarkable neutron transparency, together with their good mechanical properties at operational temperatures. Consequently, it is very important to be able to characterize the mechanical response of the irradiated cladding. The mechanical behaviour of the material can be modelled as elastoplastic with different stress-strain curves depending on the direction: radial, hoop or longitudinal direction. The ring tensile test has been proposed to determine the mechanical properties of the cladding along the hoop direction. The initial test consisted of applying a force inside the tube, by means of two half cylinders. Later Arsene and Bai [1,2] modified the experimental device to avoid tube bending at the beginning of the test. The same authors proposed a numerical method to obtain the stress-strain curve in the hoop direction from the experimental load versus displacement results and a given friction coefficient between the loading pieces and the sample [3]. This method has been used by different authors [4] with slight modifications. It is based on the existence of two universal curves under small strain hypothesis: the first correlating the hoop strain and the displacement of the loading piece and the second one correlating the hoop stress and the applied load. In this work, a new method to determine the mechanical properties of the cladding from the ring tensile test results is proposed. Non-linear geometry is considered and an iterative procedure is proposed so universal curves are not needed. A stress-strain curve is determined by combining numerical calculations with experimental results in a convergent loop. The two universal curves proposed by Arsene and Bai [3] are substituted by two relationships, one between the equivalent plastic strain in the centre of the specimen ligament and the

  16. New cladding materials and evolution of nuclear fuel components for PWR; Nouveaux materiaux de gainage et evolution des produits de combustible REP

    Energy Technology Data Exchange (ETDEWEB)

    Aubry, S. [Electricite de France (EDF), EDF Div. Combustible Nucleaire, 92 - Clamart (France); Francillon, E. [FRAMATOME ANP, Secteur Combustible, 92 - Paris-La-Defence (France); Guillet, J.L. [CEA Saclay, Dir. du Soutien Nucleaire Industriel, 91 - Gif-sur-Yvette (France)

    2004-07-01

    This paper presents recent improvements in the field of nuclear fuels made by Framatome-ANP. The first one is the use of the M5 (trade mark) alloy for the fuel cladding and guide tubes. This alloys is composed of zirconium, niobium and oxygen, it follows an optimized industrial fabrication process, it can bear combustion rates over 70 GWd/t even in harsh conditions and is strongly resistant to corrosion. Other improvements have been made in the design of the fuel assembly structure, it concerns the lower part of the one-piece tube guide for control rods and the bi-grid device whose purpose is to hold better the fuel assembly in order to reduce the fretting wear on the lower part of fuel rods. Another improvement is the doping of fuel pellets with chromium that allows, combined with an optimized micro-structure, the reduction of the volume of the gaseous fission products released in the fuel. (A.C.)

  17. ODS Ferritic/martensitic alloys for Sodium Fast Reactor fuel pin cladding

    Science.gov (United States)

    Dubuisson, Philippe; Carlan, Yann de; Garat, Véronique; Blat, Martine

    2012-09-01

    The development of ODS materials for the cladding for Sodium Fast Reactors is a key issue to achieve the objectives required for the GEN IV reactors. CEA, AREVA and EDF have launched in 2007 an important program to determine the optimal fabrication parameters, and to measure and understand the microstructure and properties before, under and after irradiation of such cladding materials. The aim of this paper is to present the French program and the major results obtained recently at CEA on Fe-9/14/18Cr1WTiY2O3 ferritic/martensitic ODS materials. The first step of the program was to consolidate Fe-9/14/18Cr ODS materials as plates and bars to study the microstructure and the mechanical properties of the new alloys. The second step consists in producing tubes at a geometry representative of the cladding of new Sodium Fast Reactors. The optimization of the fabrication route at the laboratory scale is conducted and different tubes were produced. Their microstructure depends on the martensitic (Fe-9Cr) or ferritic (Fe-14Cr) structure. To join the plug to the tube, the reference process is the welding resistance. A specific approach is developed to model the process and support the development of the welds performed within the "SOPRANO" facility. The development at CEA of Fe-9/14/18Cr new ODS materials for the cladding for GENIV Sodium Fast Reactors is in progress. The first microstructural and mechanical characterizations are very encouraging and the full assessment and qualification of this new alloys and products will pass through the irradiation of specimens, tubes, fuel pins and subassemblies up to high doses.

  18. On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim; Sanchez Espinoza, Victor Hugo [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology

    2012-11-15

    This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment. (orig.)

  19. Fuel cell assembly with electrolyte transport

    Science.gov (United States)

    Chi, Chang V.

    1983-01-01

    A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

  20. Evaluation of zinc addition on fuel cladding corrosion at the Halden test reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kolstad, E.; Symons, W.J.; Bryhn-Integrigtsen, K.; Oberlaender, B.C.

    1996-08-01

    Experimental studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor. These tests were carried out in a PWR rig inserted in the Halden reactor core. The rig simulated thermal hydraulic and coolant conditions typical of a MR. It had two flow channels where the fuel rod segments were exposed to the coolant under irradiation flux. Selected pre-characterized rodlets with fresh and pre-irradiated standard and low-tin Zircaloy-4 material were irradiated for three cycles. First cycle lasted for 110 effective full power days (EFPDs), the second for 95 EFPDs and the last 62 EFPDs. The cladding corrosion behavior was monitored by initial, interim and final oxide thickness measurements by eddy current lift-off probe. Crud sampling was performed in both channels after cycle 1 and 2. Destructive post-irradiation examinations (PIE) of two rodlets, irradiated during cycle 1 and 2, have also been completed at the conclusion of the in-pile testing. This report presents the results on oxide thickness measurements, irradiation history and water chemistry data, and the PIE.

  1. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Directory of Open Access Journals (Sweden)

    Waseem

    2016-01-01

    less than the design stress limit of the materials used for the grid [ASTM, Standard specification for precipitation hardening nickel alloy (UNSN07718 plate, sheet, and strip for high temperature service, B 670-80, USA, 2013], fuel rod [ASTM, Standard specification for wrought zirconium alloy seamless tubes for nuclear reactor fuel cladding, B 811-02, USA, 2002] and the guide thimble [ASTM, Standard specification for seamless stainless steel mechanical tubing, A 511-04, USA, 2004]. Therefore, the structural integrity criterion of CHASNUPP-1 fuel assembly is fulfilled safely at the specified tensile load.

  2. Membrane electrode assembly for a fuel cell

    Science.gov (United States)

    Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

    2006-01-01

    A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

  3. Reconstitution of fuel assemblies and core components

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, Wolfgang; Langenberger, Jan [AREVA NP GmbH (Germany)

    2012-11-01

    Due to AREVA's experience and big portfolio of techniques, reconstitution of fuel assemblies and core components at light water reactors is possible within a reasonable timeframe and with interesting cost benefit. Customer feedback indicates the sustainability of such reconstitutions. As a result, a long-term maintenance of value can be assured and early waste disposal can be avoided. (orig.)

  4. On the relative role of processes whose sequence results in crack growth in the cladding of LMFBR fuel pins

    Science.gov (United States)

    Mikhlin, E. Ya.

    1991-08-01

    Processes are discussed the joint effect of which results in crack development in austenitic steel-clad oxide fuel pins. Such processes include generation of Te which is considered as the main embrittling agent, its transport and accumulation at the cladding inner surface, where together with Cs it forms a liquid surface-acting medium, and finally, development of intergranular cracks in the cladding caused by the contact with this medium. As the process of crack growth in itself proceeds faster than accumulation of liquid surfactants at the cladding, the cracks will be able to reach the critical length only after the necessary amount of Te has been accumulated. Its accumulation is determined and therefore, controlled by the process of Te transport in the fuel grains. It is shown that the main contribution to the accumulation of Te at the cladding surface is provided by the hottest internal zones of the fuel pellet. On the basis of the analysis given, means are discussed, for inhibiting or blocking the crack growth.

  5. Cold Spray Coating Technique with FeCrAl Alloy Powder for Developing Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    Various approaches to enhance safety have been suggested, replacing current Zr-based alloys for fuel cladding with advanced materials exhibiting lower oxidation rates can be a basic solution. Many advanced materials such as FeCrAl alloys; Mn+1AXn, (MAX) phases, where n = 1 to 3, M is an early transition metal, A is an A-group (mostly IIIA and IVA, or groups 13 and 14) element and X is either carbon or nitrogen; Mo; and SiC are being considered as possible candidates. Among the proposed fuel cladding substitutes, Fe-based alloys are one of the most promising candidates owing to their excellent formability, high strength, and oxidation resistance at high temperature. In this work, the ATF technology concept of Fe-based alloy coating on the existing Zr-alloy cladding was considered and results on the optimization study for fabrication of coated tube samples were described. Result obtained from high temperature oxidation test under steam environment at 1200 .deg. C indicates that FeCrAl alloy coated Zr metal matrix may maintain its integrity during LOCA. This means that accident tolerance of FeCrAl alloy coated Zr cladding sample had been greatly improved compared to that of existing Zr-based alloy fuel cladding.

  6. Increased local corrosion of SVEA-96 fuel assemblies in KKL. Final report; Erhoehte lokale Korrosion von SVEA-96-Brennelementen im KKL. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-11-15

    In February 1997 it was noted that the cladding surface below the spacers of SVEA-96 fuel assemblies showed increased local corrosion. This phenomenon called 'Enhanced Spacer Shadow Corrosion' (ESSC) by the fuel supplier was carefully monitored during the following 4 years. Several measures have been taken in order to counteract this ESSC. In this report of the Swiss Federal Agency for the Safety of Nuclear Installations (HSK) a summary is given of the technical and licensing aspects of ESSC. Although the fundamental mechanisms for the occurrence of ESSC are not yet sufficiently understood, short-term modification to water chemistry and the increasing use of improved cladding materials have effectively reduced this phenomenon. For the justification of the use of ESSC-damaged SVEA-96 fuel assemblies, HSK established temporary criteria which are based on technical investigations by the fuel assembly supplier. Among these, a special mention can be made of the more restrictive thermo-mechanical operation limit (TMOL) curve. As proof with respect of the HSK criteria, the plant operator conducted extended inspections on fuel assemblies during serving periods in 1997-2001 (measurement of oxide thickness). The conservative aspect of the measurements was assured through destructive examinations carried out at the Hot Laboratory of the Paul Scherrer Institute (PSI). Based on the modified water chemistry and the design of the core loading for cycle 18 (2001/2002) which contains only ESSC resistant cladding materials (LK2+, LK3), the original licence basis concerning the tolerable oxide thickness on the cladding could be guarantied. This has been verified by the results of a fuel assembly examination in August 2001. Therefore, the problem of the increased corrosion of the cladding of the SVEA-96 fuel assemblies is considered as being solved

  7. 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.

  8. Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

    Energy Technology Data Exchange (ETDEWEB)

    Borring, J.; Gundtoft, H.E.; Borum, K.K.; Toft, P. [Riso National Lab. (Denmark)

    1997-08-01

    In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptance of a thinner nominal cladding than normally used today.

  9. 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.

  10. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. the combinatorial optimization problem of grouping fuel rods into a given number of rod groups with the same enrichment, and the problem of determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by the linear combination C{sub 1}X + C{sub 2}X{sub G}, where X and X{sub G} stand, respectively, for control variables giving constraint to the local power peaking factor and the gadolinium rod power. C{sub 1} and C{sub 2} are user-definable weighting factors to accommodate design preferences. The algorithm for solving this combinatorial optimization problem starts by finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering. This latter problem is solved using the method of approximation programming. A practical application is shown for a contemporary 8 x 8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  11. The reliability of untempered end plug welds on HT9-clad IFR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, D C; Porter, D L

    1987-02-01

    Welding generally leaves residual stresses in transformed weld zones, which can initiate cracks from flaws already present in the weld zones. When HT9 cools from welding temperatures, a martensite phase forms in the weld fusion zone and heat-affected zone. Because this martensite phase is hard and brittle, it is particularly susceptible to cracking aggravated by residual stresses. This causes concern over the use of untempered welds on HT9-clad fuel elements. To determine if residual stresses present in end-plug weld zones would affect fuel pin performance, HT9 capsules with prototypic TIG- and CD-welded end plugs (in the tempered and as-welded conditions) were pressurized to failure at room temperature, 550{sup 0}C, and 600{sup 0}C. None of the capsules failed in a weld zone. To determine the effects of reactor operating temperatures on untempered welds, prototypic TIG welds were tempered at reactor bulk sodium temperature and an expected sodium outlet temperature for various lengths of time. Subsequent tensile and burst tests of these specimens proved that any embrittling effects that may have been induced in these welds were of no consequence. Hardness tests on longitudinal sections of welds indicated the amount of tempering a weld will receive inreactor after relatively short lengths of time. The pressure burst tests proved that untemperted welds on HT9-clad fuel elements are as reliable as tempered welds; any residual stresses in untempered weld zones were of no consequence. The tempering test showed that welds used in the as-welded condition will sufficiently temper in 7 days at 550{sup 0}C, but will not, sufficiently temper in 7 days at bulk sodium temperature. A comparison of the structure of laser welds to those of CD and TIG welds indicated that untempered laser welds will perform and temper in a manner similar to the TIG welds tested in this effort.

  12. Corrosion of the AlFeNi alloy used for the fuel cladding in the Jules Horowitz research reactor

    Science.gov (United States)

    Wintergerst, M.; Dacheux, N.; Datcharry, F.; Herms, E.; Kapusta, B.

    2009-09-01

    The AlFeNi aluminium alloy (1 wt% Fe, 1 wt% Ni, 1 wt% Mg) is expected to be used as nuclear fuel cladding for the Jules Horowitz experimental reactor. To guarantee a safe behaviour of the fuel, a good understanding of the fuel clad corrosion mechanisms is required. In this field, the experimental characterization of the selected alloy was performed. Then experimental studies of the aluminium alloy corrosion product obtained in autoclaves have shown an oxide film composed of two layers. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion parallel to a dissolution-precipitation process to form the outer zone. Dynamic experiments at 70 °C have demonstrated that a solid diffusion step controls the release kinetic. 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.

  13. Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

    Science.gov (United States)

    Rico, A.; Martin-Rengel, M. A.; Ruiz-Hervias, J.; Rodriguez, J.; Gomez-Sanchez, F. J.

    2014-09-01

    It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young's modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young's modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found.

  14. Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Rico, A., E-mail: alvaro.rico@urjc.es [DIMME, Departamento de Tecnología Mecánica, Universidad Rey Juan Carlos, c/Tulipán s/n, E-28933 Móstoles, Madrid (Spain); Martin-Rengel, M.A., E-mail: mamartin@mater.upm.es [Departamento de Ciencia de los Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren SN, E-28040 Madrid (Spain); Ruiz-Hervias, J., E-mail: jesus.ruiz@upm.es [Departamento de Ciencia de los Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren SN, E-28040 Madrid (Spain); Rodriguez, J. [DIMME, Departamento de Tecnología Mecánica, Universidad Rey Juan Carlos, c/Tulipán s/n, E-28933 Móstoles, Madrid (Spain); Gomez-Sanchez, F.J., E-mail: javier.gomez@amsimulation.com [Advanced Material Simulation, S.L, Madrid (Spain)

    2014-09-15

    It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found.

  15. Obtention of the constitutive equation of hydride blisters in fuel cladding from nanoindentation tests

    Science.gov (United States)

    Martin Rengel, M. A.; Gomez, F. J.; Rico, A.; Ruiz-Hervias, J.; Rodriguez, J.

    2017-04-01

    It is well known that the presence of hydrides in nuclear fuel cladding may reduce its mechanical and fracture properties. This situation may be worsened as a consequence of the formation of hydride blisters. These blisters are zones with an extremely high hydrogen concentration and they are usually associated to the oxide spalling which may occur at the outer surface of the cladding. In this work, a method which allows us to reproduce, in a reliable way, hydride blisters in the laboratory has been devised. Depth-sensing indentation tests with a spherical indenter were conducted on a hydride blister produced in the laboratory with the aim of measuring its mechanical behaviour. The plastic stress-strain curve of the hydride blister was calculated for first time by combining depth-sensing indentation tests results with an iterative algorithm using finite element simulations. The algorithm employed reduces, in each iteration, the differences between the numerical and the experimental results by modifying the stress-strain curve. In this way, an almost perfect adjustment of the experimental data was achieved after several iterations. The calculation of the constitutive equation of the blister from nanoindentation tests, may involve a lack of uniqueness. To evaluate it, a method based on the optimization of parameters of analytical equations has been proposed in this paper. An estimation of the error which involves this method is also provided.

  16. Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

    Science.gov (United States)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  17. In-core measurements of fuel-clad interactions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett Peter

    2008-10-15

    A combination of on-line measurement techniques was used to demonstrate AOA in the Halden reactor: - Diameter gauge to demonstrate crud deposition - Coolant flow and temperature measurements to show effect of crud on thermal-hydraulic conditions - Neutron detectors to show power depression caused by boron in the crud. - Coolant chemistry analyses provided supporting evidence of AOA - Lithium return during shutdown - PIE showed that the type of crud observed in US plants suffering severe AOA can be reproduced. Loop systems allow testing under LWR thermal-hydraulic and water chemistry conditions. - A combination of on-line instrumentation allows measurements of complicated phenomena, egPWR AOA. - Techniques are under development to allow on-line measurements of fuel clad corrosion

  18. Microstructure stability of candidate stainless steels for Gen-IV SCWR fuel cladding application

    Science.gov (United States)

    Li, Jian; Zheng, W.; Penttilä, S.; Liu, P.; Woo, O. T.; Guzonas, D.

    2014-11-01

    In the past few years, significant progress has been made in materials selection for Gen-IV SCWR fuel cladding applications. Current studies indicate that austenite stainless steels such as 310H are promising candidates for in-core applications. Alloys in this group are promising for their corrosion resistance, SCC resistance, high temperature mechanical properties and creep resistance at temperatures up to 700 °C. However, one under-studied area of this alloy is the long-term microstructure stability under the proposed reactor operating condition. Unstable microstructure not only results in embrittlement but also has the potential to reduce their resistance to corrosion or stress-corrosion cracking. In this study, stainless steels 310H and 304H were tested for their SCWR corrosion resistance and microstructure stability.

  19. Research on Measuring Technology for In-pile Fuel Element Testing

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The tested fuel assembly for In-pile test for PWR fuel element with instrumentation consisted of 4instrumented fuel elements and total 12 sets of transducers. Double claddings are adopted to raise fueltemperature. Two fuel elements each have 2 thermocouples for measuring separately the fuel centerlinetemperature and the cladding surface temperature. The other two elements have membrane type oressure

  20. Past research and fabrication conducted at SCK•CEN on ferritic ODS alloys used as cladding for FBR's fuel pins

    Science.gov (United States)

    De Bremaecker, Anne

    2012-09-01

    -destructive tests (ultrasonic and eddy currents) were also developed. In-pile creep in argon and in liquid sodium was deeply studied on pressurized segments irradiated up to 75 dpaNRT. Finally two fuel assemblies cladded with such ODS alloys were irradiated in Phenix to the max dose of 90 dpa. Creep deformation and swelling were limited but the irradiation-induced embrittlement became acute. The programme was stopped shortly after the Chernobyl disaster, before the embrittlement problem was solved.

  1. ATR LEU Monolithic Foil-Type Fuel with Integral Cladding Burnable Absorber – Neutronics Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Gray Chang

    2012-03-01

    The Advanced Test Reactor (ATR), currently operating in the United States, is used for material testing at very high neutron fluxes. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting HEU driven reactor cores to low-enriched uranium (LEU) cores. The burnable absorber - 10B, was added in the inner and outer plates to reduce the initial excess reactivity, and to improve the peak ratio of the inner/outer heat flux. The present work investigates the LEU Monolithic foil-type fuel with 10B Integral Cladding Burnable Absorber (ICBA) design and evaluates the subsequent neutronics operating effects of this proposed fuel designs. The proposed LEU fuel specification in this work is directly related to both the RERTR LEU Development Program and the Advanced Test Reactor (ATR) LEU Conversion Project at Idaho National Laboratory (INL).

  2. Advanced Steels for Accident Tolerant Fuel Cladding in Current Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    After the March 2011 Fukushima events, the U.S. Congress directed the Department of Energy (DOE) to focus efforts on the development of fuel cladding materials with enhanced accident tolerance. In comparison with the stand-ard UO2-Zirconium based system, the new fuels need to tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operation conditions. Advanced steels such as iron-chromium-aluminum (FeCrAl) alloys are being investigated for degradation behavior both under normal operation conditions in high temperature water (e.g. 288°C) and under accident conditions for reaction with steam up to 1400°C. Commercial and experimental alloys were tested for several periods of time in 100% superheated steam from 800°C to 1475°C. Results show that FeCrAl alloys significantly outperform the resistance in steam of the current zirconium alloys.

  3. Formation of intermetallic compound at interface between rare earth elements and ferritic-martensitic steel by fuel cladding chemical interaction

    Institute of Scientific and Technical Information of China (English)

    Jun Hwan Kim; Byoung Oon Lee; Chan Bock Lee; Seung Hyun Jee; Young Soo Yoon

    2012-01-01

    The intermetallic compounds formation at interface between rare earth elements and clad material were investigated to demonstrate the effects of rare earth elements on fuel-cladding chemical interaction (FCCI) behavior.Mischmetal (70Ce-30La) and Nd were prepared as rare earth elements.Diffusion couple testing was performed on the rare earth elements and cladding (9Cr2W steel) near the operation temperature of(sodium-cooled fast reactor) SFR fuel.The performance of a diffusion barrier consisting of Zr and V metallic foil against the rare earth elements was also evaluated.Our results showed that Ce and Nd in the rare earth elements and Fe in the clad material interdiffused and reacted to form intermetallic species according to the parabolic rate law,describing the migration of the rare earth element.The diffusion of Fe limited the reaction progress such that the entire process was governed by the cubic rate law.Rare earth materials could be used as a surrogate for high burnup metallic fuels,and the performance of the barrier material was demonstrated to be effective.

  4. 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)

  5. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  6. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO2 or ZrO2. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

  7. Impact of the use of the ferritic/martensitic ODS steels cladding on the fuel reprocessing PUREX process

    Science.gov (United States)

    Gwinner, B.; Auroy, M.; Mas, D.; Saint-Jevin, A.; Pasquier-Tilliette, S.

    2012-09-01

    Some ferritic/martensitic oxide dispersed strengthened (F/M ODS) steels are presently developed at CEA for the fuel cladding of the next generation of sodium fast nuclear reactors. The objective of this work is to study if this change of cladding could have any consequences on the spent fuel reprocessing PUREX process. During the fuel dissolution stage the cladding can actually be corroded by nitric acid. But some process specifications impose not to exceed a limit concentration of the corrosion products such as iron and chromium in the dissolution medium. For that purpose the corrosion behavior of these F/M ODS steels is studied in hot and concentrated nitric acid. The influence of some metallurgical parameters such as the chromium content, the elaboration process and the presence of the yttrium oxides is first discussed. The influence of environmental parameters such as the nitric acid concentration, the temperature and the presence of oxidizing species coming from the fuel is then analyzed. The corrosion rate is characterized by mass loss measurements and electrochemical tests. Analyses of the corroded surface are carried out by X-ray photoelectron spectroscopy.

  8. Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations

    Energy Technology Data Exchange (ETDEWEB)

    Greiner, Miles [Univ. of Nevada, Reno, NV (United States)

    2017-03-31

    Radial hydride formation in high-burnup used fuel cladding has the potential to radically reduce its ductility and suitability for long-term storage and eventual transport. To avoid this formation, the maximum post-reactor temperature must remain sufficiently low to limit the cladding hoop stress, and so that hydrogen from the existing circumferential hydrides will not dissolve and become available to re-precipitate into radial hydrides under the slow cooling conditions during drying, transfer and early dry-cask storage. The objective of this research is to develop and experimentallybenchmark computational fluid dynamics simulations of heat transfer in post-pool-storage drying operations, when high-burnup fuel cladding is likely to experience its highest temperature. These benchmarked tools can play a key role in evaluating dry cask storage systems for extended storage of high-burnup fuels and post-storage transportation, including fuel retrievability. The benchmarked tools will be used to aid the design of efficient drying processes, as well as estimate variations of surface temperatures as a means of inferring helium integrity inside the canister or cask. This work will be conducted effectively because the principal investigator has experience developing these types of simulations, and has constructed a test facility that can be used to benchmark them.

  9. Results of High-Temperature Heating Test for Irradiated U-10Zr(-5Ce with T92 Cladding Fuel

    Directory of Open Access Journals (Sweden)

    June-Hyung Kim

    2016-11-01

    Full Text Available A microstructure observation using an optical microscope, SEM and EPMA was performed for the irradiated U-10Zr and U-10Zr-5Ce fuel slugs with a T92 cladding specimen after a high-temperature heating test. Also, the measured eutectic penetration rate was compared with the value predicted by the existing eutectic penetration correlation being used for design and modeling purposes. The heating temperature and duration time for the U-10Zr/T92 specimen were 750 °C and 1 h, and those for the U-10Zr-5Ce/T92 specimen were 800 °C and 1 h. In the case of the U-10Zr/T92 specimen, the migration phenomena of U, Zr, Fe, and Cr as well as the Nd lanthanide fission product were observed at the eutectic melting region. The measured penetration rate was similar to the value predicted by the existing eutectic penetration rate correlation. In addition, when comparing with measured eutectic penetration rates for the unirradiated U-10Zr fuel slug with FMS (ferritic martensitic steel, HT9 or Gr.91 cladding specimens which had been reported in the literature, the measured eutectic penetration rate for the irradiated fuel specimen was higher than that for the unirradiated U-10Zr specimen. In the case of the U-10Zr-5Ce/T92 specimen in which there had been a gap between the fuel slug and cladding after the irradiation test, the eutectic melting region was not found because contact between the fuel slug and cladding did not take place during the heating test.

  10. Post-irradiation examination of AlFeNi cladded U 3Si 2 fuel plates irradiated under severe conditions

    Science.gov (United States)

    Leenaers, A.; Koonen, E.; Parthoens, Y.; Lemoine, P.; Van den Berghe, S.

    2008-04-01

    Three full size AlFeNi cladded U 3Si 2 fuel plates were irradiated in the BR2 reactor of the Belgian Nuclear Research Centre (SCK·CEN) under relatively severe, but well defined conditions. The irradiation was part of the qualification campaign for the fuel to be used in the future Jules Horowitz reactor in Cadarache, France. After the irradiation, the fuel plates were submitted to an extensive post-irradiation campaign in the hot cell laboratory of SCK·CEN. The PIE shows that the fuel plates withstood the irradiation successfully, as no detrimental defects have been found. At the cladding surface, a multilayered corrosion oxide film has formed. The U-Al-Si layer resulting from the interaction between the U 3Si 2 fuel and the Al matrix, has been quantified as U(Al,Si) 4.6. It is found that the composition of the fuel particles is not homogenous; zones of USi and U 3Si 2 are observed and measured. The fission gas-related bubbles generated in both phases show a different morphology. In the USi fuel, the bubbles are small and numerous while in U 3Si 2 the bubbles are larger but there are fewer.

  11. Critical Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion Controlled Cavity Growth

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, T.A.; Rosen, R.S.; Kassner, M.E.

    1999-12-01

    Interim dry storage of spent nuclear fuel (SNF) rods is of critical concern because a shortage of existing SNF wet storage capacity combined with delays in the availability of a permanent disposal repository has led to an increasing number of SNF rods being placed into interim dry storage. Safe interim dry storage must be maintained for a minimum of twenty years according to the Standard Review Plan for Dry Cask Storage Systems [1] and the Code of Federal Regulations, 10 CFR Part 72 [2]. Interim dry storage licensees must meet certain safety conditions when storing SNF rods to ensure that there is a ''very low probability (e.g. 0.5%) of cladding breach during long-term storage'' [1]. Commercial SNF typically consists of uranium oxide pellets surrounded by a thin cladding. The cladding is usually an {alpha}-zirconium based alloy know as ''Zircaloy''. In dry storage, the SNF rods are confined in one of several types of cask systems approved by the Nuclear Regulatory Commission (NRC). ''The cask system must be designed to prevent degradation of fuel cladding that results in a type of cladding breach, such as axial-splits or ductile fracture, where irradiated UO{sub 2} particles may be released. In addition, the fuel cladding should not degrade to the point where more than one percent of the fuel rods suffer pinhole or hairline crack type failure under normal storage conditions [1].'' The NRC has approved two models [3,4] for use by proposed dry storage licensees to determine the maximum initial temperature limit for nuclear fuel rods in dry storage that supposedly meet the above criteria and yield consistent temperature limits. Though these two models are based on the same fundamental failure theory, different assumptions have been made including the choice of values for material constants in the failure equation. This report will examine and compare the similarities and inconsistencies of these two models

  12. 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

  13. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  14. CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.

    2014-06-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  15. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  16. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-31

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  17. Galvanic corrosion of Mg-Zr fuel cladding and steel immobilized in Portland cement and geopolymer at early ages

    Science.gov (United States)

    Rooses, Adrien; Lambertin, David; Chartier, David; Frizon, Fabien

    2013-04-01

    Galvanic corrosion behaviour of Mg-Zr alloy fuel cladding and steel has been studied in Ordinary Portland cement and Na-geopolymer. Portland cements implied the worse magnesium corrosion performances due to the negative effects of cement hydrates, grinding agents and gypsum on the galvanic corrosion. Galvanic corrosion in Na-geopolymer paste remains very low. Silicates and fluoride from the geopolymer activation solution significantly improve the corrosion resistance of the magnesium alloy while coupling with a cathode.

  18. Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scott, Jeffrey E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Duffield, Andrew N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weinberg, Richard Y. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alexander, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hudson, Richard W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mihaila, Bogdan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Liu, Cheng [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lovato, Manuel L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-08-26

    Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-23

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

  20. Westinghouse Fuel Assemblies Performance after Operation in South-Ukraine NPP Mixed Core

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Kulish, G. V.; Slyeptsov, O.; Slyeptsov, S.; Aleshin, Y.; Sparrow, S.; Lashevych, P.; Sokolov, D.; Latorre, Richard

    2013-09-14

    The evaluation of WWER-1000 Westinghouse fuel performance was done using the results of post–irradiation examinations of six LTAs and the WFA reload batches that have operated normally in mixed cores at South-Ukraine NPP, Unit-3 and Unit-2. The data on WFA/LTA elongation, FR growth and bow, WFA bow and twist, RCCA drag force and drag work, RCCA drop time, FR cladding integrity as well as the visual observation of fuel assemblies obtained during the 2006-2012 outages was utilized. The analysis of the measured data showed that assembly growth, FR bow, irradiation growth, and Zr-1%Nb grid and ZIRLO cladding corrosion lies within the design limits. The RCCA drop time measured for the LTA/WFA is about 1.9 s at BOC and practically does not change at EOC. The measured WFA bow and twist, and data of drag work on RCCA insertion showed that the WFA deformation in the mixed core is mostly controlled by the distortion of Russian FAs (TVSA) having the higher lateral stiffness. The visual inspection of WFAs carried out during the 2012 outages revealed some damage to the Zr-1%Nb grid outer strap for some WFAs during the loading sequence. The performed fundamental investigations allowed identifying the root cause of grid outer strap deformation and proposing the WFA design modifications for preventing damage to SG at a 225 kg handling trip limit.

  1. Calibration of spent fuel measurement assembly

    Science.gov (United States)

    Koleska, Michal; Viererbl, Ladislav; Marek, Milan

    2014-11-01

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system.

  2. High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments

    Science.gov (United States)

    Pint, B. A.; Terrani, K. A.; Brady, M. P.; Cheng, T.; Keiser, J. R.

    2013-09-01

    Alternative fuel cladding materials to Zr alloys are being investigated for enhanced accident tolerance, which specifically involves oxidation resistance to steam or steam-H2 environments at ⩾1200 °C for short times. Based on a comparison of a range of commercial and model alloys, conventional austenitic steels do not have sufficient oxidation resistance with only ˜18Cr-10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application. Results at 1350 °C indicated that FeCrAl alloys and CVD SiC remain oxidation resistant in steam. At 1200 °C, high (⩾25% Cr) ferritic alloys appear to be good candidates for this application. Higher pressures (up to 20.7 bar) and H2 additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys, but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed for type 317 L tubing in a H2-50%H2O environment at 10.3 bar compared to 100% H2O.

  3. Investigation of silver and iodine transport through silicon carbide layers prepared for nuclear fuel element cladding

    Science.gov (United States)

    Friedland, E.; van der Berg, N. G.; Malherbe, J. B.; Hancke, J. J.; Barry, J.; Wendler, E.; Wesch, W.

    2011-03-01

    Transport of silver and iodine through polycrystalline SiC layers produced by PBMR (Pty) Ltd. for cladding of TRISO fuel kernels was investigated using Rutherford backscattering analysis and electron microscopy. Fluences of 2 × 10 16 Ag + cm -2 and 1 × 10 16 I + cm -2 were implanted at room temperature, 350 °C and 600 °C with an energy of 360 keV, producing an atomic density of approximately 1.5% at the projected ranges of about 100 nm. The broadening of the implantation profiles and the loss of diffusors through the front surface during vacuum annealing at temperatures up to 1400 °C was determined. The results for room temperature implantations point to completely different transport mechanisms for silver and iodine in highly disordered silicon carbide. However, similar results are obtained for high temperature implantations, although iodine transport is much stronger influenced by lattice defects than is the case for silver. For both diffusors transport in well annealed samples can be described by Fickian grain boundary diffusion with no abnormal loss through the surface as would be expected from the presence of nano-pores and/or micro-cracks. At 1100 °C diffusion coefficients for silver and iodine are below our detection limit of 10 -21 m 2 s -1, while they increase into the 10 -20 m 2 s -1 range at 1300 °C.

  4. Revisiting the method to obtain the mechanical properties of hydrided fuel cladding in the hoop direction

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A., E-mail: mamartin@mater.upm.es [Departamento de Ciencia de Materiales, UPM, ETSI Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain); Gomez Sanchez, F.J., E-mail: javier.gomez@amsimulation.com [Advanced Material Simulation, S.L (Spain); Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, ETSI Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2012-10-15

    The method reported in the literature to calculate the stress-strain curve of nuclear fuel cladding from ring tensile test is revisited in this paper and a new alternative is presented. In the former method, two universal curves are introduced under the assumption of small strain. In this paper it is shown that these curves are not universal, but material-dependent if geometric nonlinearity is taken into account. The new method is valid beyond small strains, takes geometric nonlinearity into consideration and does not need universal curves. The stress-strain curves in the hoop direction are determined by combining numerical calculations with experimental results in a convergent loop. To this end, ring tensile tests were performed in unirradiated hydrogen-charged samples. The agreement among the simulations and the experimental results is excellent for the range of concentrations tested (up to 2000 wppm hydrogen). The calculated stress-strain curves show that the mechanical properties do not depend strongly on the hydrogen concentration, and that no noticeable strain hardening occurs. However, ductility decreases with the hydrogen concentration, especially beyond 500 wppm hydrogen. The fractographic results indicate that as-received samples fail in a ductile fashion, whereas quasicleavage is observed in the hydrogen-charged samples.

  5. Characterization of Zircaloy-4 tubing procured for fuel cladding research programs

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, R.H. (comp.)

    1976-06-14

    A quantity of Zircaloy-4 tubing (10.92 mm outside diameter by 0.635 mm wall thickness) was purchased specifically for use in a number of related fuel cladding research programs sponsored by the Division of Reactor Safety Research, Nuclear Regulatory Commission (NRC/RSR). Identical tubing (produced simultaneously and from the same ingot) was purchased concurrently by the Electric Power Research Institute (EPRI) for use in similar research programs sponsored by that organization. In this way, source variability and prior fabrication history were eliminated as parameters, thus permitting direct comparison (as far as as-received material properties are concerned) of experimental results from the different programs. The tubing is representative of current reactor technology. Consecutive serial numbers assigned to each tube identify the sequence of the individual tubes through the final tube wall reduction operation. The report presented documents the procurement activities, provides a convenient reference source of manufacturer's data and tubing distribution to the various users, and presents some preliminary characterization data. The latter have been obtained routinely in various research programs and are not complete. Although the number of analyses, tests, and/or examinations performed to date are insufficient to draw statistically valid conclusions with regard to material characterization, the data are expected to be representative of the as-received tubing. It is anticipated that additional characterizations will be performed and reported routinely by the various research programs that use the tubing.

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Vaibhaw, Kumar [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)], E-mail: krvaibhaw@yahoo.co.in; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N. [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)

    2008-12-15

    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 ({approx}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{sub 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.

  8. Model of fracture for the Zry cladding of nuclear fuel rods included in the code DIONISIO 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: soba@cnea.gov.ar; Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-12-15

    The DIONISIO code describes most of the main phenomena occurring in a fuel rod during normal operation of a nuclear power reactor. Starting from the irradiation history, the code predicts the temperature distribution, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the internal rod volume, gas mixing, pressure increase, irradiation growth of the cladding, development of an oxide layer on its surface and hydrogen uptake, restructuring and grain growth in the pellet. This work presents the model of Zircaloy fracture included in the code DIONISIO 1.0. The model of pellet-cladding mechanical interaction (PCMI) provides the forces caused by the solid-solid contact which add to the changing internal pressure and to the constant external pressure. Besides, the program evaluates the effects of a corrosive atmosphere (stress corrosion cracking, SCC) internal or external. With these data, the code calculates the J integral around the tip of an initiated crack, and proceeds to analyze, according to the quantity of corrosive substance dissolved and the cladding stress field, if the crack remains unchanged, if it grows due to the I-SCC mechanism, or if propagation is ductile, following the R curve of the material. Results corresponding to different PHWR and PWR reactors are presented and compared with code results. In particular, good agreement is obtained in the simulation of MOX experiments, where the cladding failed due to propagation of cracks originated in SCC.

  9. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Science.gov (United States)

    Jacobsson Svärd, Staffan; Holcombe, Scott; Grape, Sophie

    2015-05-01

    A fuel assembly operated in a nuclear power plant typically contains 100-300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which

  10. Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John

    2016-10-01

    Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

  11. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  12. 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.

  13. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  14. 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.

  15. Modelling and modal properties of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2011-12-01

    Full Text Available The paper deals with the modelling and modal analysis of the hexagonal type nuclear fuel assembly. This very complicated mechanical system is created from the many beam type components shaped into spacer grids. The cyclic and central symmetry of the fuel rod package and load-bearing skeleton is advantageous for the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and skeleton linked by several spacer grids in horizontal planes. The derived mathematical model is used for the modal analysis of the Russian TVSA-T fuel assembly and validated in terms of experimentally determined natural frequencies, modes and static deformations caused by lateral force and torsional couple of forces. The presented model is the first necessary step for modelling of the nuclear fuel assembly vibration caused by different sources of excitation during the nuclear reactor VVER type operation.

  16. Carbon 14 distribution in irradiated BWR fuel cladding and released carbon 14 after aqueous immersion of 6.5 years

    Energy Technology Data Exchange (ETDEWEB)

    Sakuragi, T. [Radioactive Waste Management Funding and Research Center, Tsukishima 1-15-7, Chuo City, Tokyo, 104-0052 (Japan); Yamashita, Y.; Akagi, M.; Takahashi, R. [TOSHIBA Corporation, Ukishima Cho 4-1, Kawasaki Ward, Kawasaki, 210-0862 (Japan)

    2016-07-01

    Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 ({sup 14}C), which is mainly generated by the {sup 14}N(n,p){sup 14}C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF). In the present work, specific activity of {sup 14}C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The {sup 14}C specific activity of the base metal was 1.49*10{sup 4} Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the {sup 17}O(n,α){sup 14}C reaction, the {sup 14}C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, {sup 14}C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the {sup 14}C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released {sup 14}C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the {sup 14}C abundance and the low leaching rate

  17. In-situ Observation of Boiling Dynamics on Fuel Cladding Surface in Non-pressurized Water Using Acoustic Emission Method

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Kaige; Baek, Seung Heon; Shim, Hee-Sang; Hur, Do Haeng; Lee, Deok Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In the PWR primary coolant system, a phenomenon of axial offset anomaly (AOA) can be caused due to accumulated boron hide out in porous CRUD deposition on the fuel cladding surface. Up to now, the CRUD deposition has been well known to be driven by subcooled nucleate boiling (SNB) on the cladding surface based on large scale experimental work. Therefore, monitoring and evaluation of the SNB-phenomenon is an important approach to study the CRUD deposition. Many attempts have been made to study the SNB and CRUD deposition using thermal hydraulic or model calculation. However, a comprehensive understanding of the SNB during CRUD deposition is still far from being realized. Acoustic emission (AE) technique, as an in-situ nondestructive evaluation (NDE) method, has been widely used to monitor the boiling activity in containers and pipes. Accordingly, this work aimed to investigate the exact AE characteristics of SNB-phenomenon on the fuel cladding surface at atmospheric pressure, with the purpose of providing an experimental groundwork for the AE investigation on SNB in high-temperature pressurized coolant system. In this study, we conducted an in-situ experimental observation of the bubble dynamic of SNB in non-pressurized water at atmospheric pressure using AE method. The AE of heater noise was confirmed to cluster between 8 and 26 khz. Three AE groups were detected during the boiling process in the Snob zones. AE group 1 and 3 seemed to be the results of bubble growth and collapse, while bubble departure from the cladding surface was reasonably associated with an isolated AE group 2.

  18. Dynamic response of nuclear fuel assembly excited by pressure pulsations

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2012-12-01

    Full Text Available The paper deals with dynamic load calculation of the hexagonal type nuclear fuel assembly caused by spatial motion of the support plates in the reactor core. The support plate motion is excited by pressure pulsations generated by main circulation pumps in the coolant loops of the primary circuit of the nuclear power plant. Slightly different pumps revolutions generate the beat vibrations which causes an amplification of fuel assembly component dynamic deformations and fuel rods coating abrasion. The cyclic and central symmetry of the fuel assembly makes it possible the system decomposition into six identical revolved fuel rod segments which are linked with central tube and skeleton by several spacer grids in horizontal planes.The modal synthesis method with condensation of the fuel rod segments is used for calculation of the normal and friction forces transmitted between fuel rods and spacer grids cells.

  19. FBR core design with the composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, M.W.

    Although calculations are preliminary, overall feasibility of an FBR core design with the composite fuel assembly has been demonstrated. The advantaged over the heterogeneous design is that large variances in assembly mixed mean outlet temperatures are eliminated. Also, the effective enrichment of an assembly may easily be adjusted by varying the number of fertile pins per assembly, thus making it possible to flatten the core radial power profile. The use of the composite fuel assembly may in the future offer a significant alternative to heterogeneous FBR core design.

  20. Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

    Energy Technology Data Exchange (ETDEWEB)

    Jo, HangJin; Kim, Jin Man [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Yeom, Hwasung [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States); Lee, Gi Cheol [Department of Mechanical Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Kiyofumi, Moriyama; Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Sridharan, Kumar; Corradini, Michael [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States)

    2015-09-15

    Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding.

  1. Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program). [LWBR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, R.C.; Sherman, J.

    1978-11-01

    Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability.

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

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B. Jr.

    1999-10-21

    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.

  3. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  4. Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

    Science.gov (United States)

    Jung, Ju Ang; Kim, Seung Hyun; Shin, Sang Hun; Bang, In Cheol; Kim, Ji Hyun

    2013-09-01

    As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.

  5. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  6. Effect of the Boron and Nitrogen on precipitation behavior in modified 9Cr steel for SFR fuel cladding after aging

    Energy Technology Data Exchange (ETDEWEB)

    Jeog, Eun-Hee; Kim, Young Do [Hanyang University, Seoul (Korea, Republic of); Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances against void swelling. Because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels are preferable to utilize in the fuel cladding of an SFR in KAERI. The soluble boron reduces the coarsening rate of M{sub 23}C{sub 6} carbides along boundaries near prior austenite grain boundaries during creep, enhancing the boundary and sub-boundary hardening for up to long times. The enhancement of boundary and sub-boundary hardening retards the onset of acceleration creep, which decreases the minimum creep rate and improves the creep life. It has been reported that the excess addition of boron and nitrogen promotes the formation of boron nitrides during normalizing heat treatment, which significantly reduces soluble B and N concentrations and offsets the benefit due to boron and nitrogen. In this study, comparison of the microstructure and mechanical properties on SFR fuel cladding steel with different B and N contents after aging were carried out. The addition of B stabilizes the M{sub 23}C{sub 6}, hence the coarsening of M{sub 23}C{sub 6} was not observed in alloy 1 after 7000 hours aging. The size distribution of an alloy 2 was not largely changed with aging time, and this phenomena would be caused by an addition of nitrogen, by stabilize the nitride precipitates such as MX and M{sub 2}X.

  7. Physical and Numerical Difficulties in Computer Modelling of Pellet-Cladding Contact Problems for Burned-Up Fuel

    Directory of Open Access Journals (Sweden)

    M. Dostál

    2005-01-01

    Full Text Available The importance of fuel reliability is growing due to the deregulated electricity market and the demands on operability and availability to the electricity grid of nuclear units. Under these conditions of fuel exploitation, the problems of PCMI (Pellet-Cladding Mechanical Interaction are very important from the point of view of fuel rod integrity and reliability. Severe loading is thermophysically and mechanically expressed as a greater probability of cladding failure especially during power maneuvering. We have to be able to make a realistic prediction of safety margins, which is very difficult by using computer simulation methods. NRI (Nuclear Research Institute has recently been engaged in developing 2D and 3D FEM (Finite Element Method based models dealing with this problem. The latest effort in this field has been to validate 2D r-z models developed in the COSMOS/M system against calculations using the FEMAXI-V code. This paper presents a preliminary comparison between classical FEM based integral code calculations and new models that are still under development. The problem has not been definitely solved. The presented data is of a preliminary nature, and several difficult problems remain to be solved. 

  8. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    Science.gov (United States)

    Gussev, M. N.; Byun, T. S.; Yamamoto, Y.; Maloy, S. A.; Terrani, K. A.

    2015-11-01

    One of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filtering unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.

  9. 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

  10. Proficiency Testing Schemes of a Fuel Assembly Performance Method by Comparing of Measurement Results of Two Tester

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Kim, J. Y.; Lee, K. H.; Lee, Y. H.; Kim, H. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    The purpose of this work is to establish the proficiency testing schemes of a fuel assembly for nonstandard test method case. As the nuclear regulatory guide, 'the testing and inspections to be performed to verify the design characteristics of the fuel system components, including clad integrity, dimensions, fuel enrichment, burnable poison concentration, absorber composition, and characteristics of the fuel, absorber, and poison pellets, should be described. {approx}'. In this guide, the fuel assembly test method is as that, the lateral and axial stiffness, lateral vibration, lateral and axial impact and the rotational stiffness test. These method cases are very important for the license service and providing some input data for the accident analysis model of FA. Therefore, all of these tests have to be executed as the authorized standard, for example, Korea Laboratory Accreditation Scheme (KOLAS). Unfortunately, the performance tests of a FA did not certified by the KOLAS. In order to receive the authorized test scheme, the proficiency testing schemes is most important item. For non-standard test case, the most of these tests be normally executed through the inter-laboratory comparisons. However, there is no standard, no certified reference material (CRM) for pressurized water reactor (PWR) fuel assembly. In this case, the most important point is that how to verify the validity of the performance test method of a fuel. Therefore, the inter-personnel testing scheme is proposed for this. For the proficiency testing of a fuel assembly performance test, the lateral bending test of a fuel assembly (FA) is executed using FAMeCT. The FAMeCT is a tester of a versatile function for a mechanical characterization of an actual size FA. Because of the absence of the CRM, the t-test method was selected. Null and alternative hypotheses were assumed and then t-value was evaluated as these hypotheses.

  11. Ultrasonic Inspection for Zirconium Alloy Nuclear Fuel Cladding Tubes%核燃料锆合金包壳管的超声波探伤

    Institute of Scientific and Technical Information of China (English)

    夏健文; 韩承

    2016-01-01

    介绍压水堆核燃料锆合金包壳管(Φ10.0 mm×0.70 mm)的超声波自动探伤方法和工艺,讨论不同长度、宽度、深度、角度的纵向和横向人工缺陷的超声响应结果.通过对检测出缺陷的典型包壳管进行金相解剖,确定缺陷性质和实际尺寸,验证超声探伤结果.针对实际探伤中的问题,考虑质量和成本控制,提出对不同缺陷的验收准则.实践应用表明,现行探伤方法和工艺能检出管材不同位置处10μm级的微小缺陷.但受缺陷的类型、取向的影响,探伤仪检测得到的回波幅度并不能完全真实地反应缺陷的实际大小和性质,需要在实际探伤时针对管材的制造工艺水平采取适当的加严措施,对不同的缺陷加以控制,才能更好地保证核燃料包壳管的质量.%The cladding tube is the main component of the nuclear fuel assembly,and as the first protective barrier,its quality is very important for the safe operation of nuclear power plants.After the completion of cladding tubes,a non-destructive testing is required,in which the ultrasonic inspection is a primary method.This paper introduces the ultrasonic flaw testing method and techniques of the zirconium alloy nuclear fuel cladding tubes for pressurized water reactor (PWR),which used in automatic ultrasonic inspection equipment,and discusses the detector response to the longitudinal and transverse artificial defects of different length,width,depth and angle.Its actual shape and size are measured by metallographic anatomical analysis for some typical defects to confirm the flaw detection results.Consider its quality and cost control,the acceptance rules are proposed for different defects.The application shows that the existing detection method and process can inspect the fine defects about 10μm at different locations of the cladding tube.Due to the influence of the defect type and orientation,the echo amplitude obtained by the detector is not completely true to the

  12. Advanced ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N [ORNL; Unocic, Kinga A [ORNL; Hoelzer, David T [ORNL; Pint, Bruce A [ORNL

    2014-09-01

    ODS FeCrAl alloys are being developed with optimum composition and properties for accident tolerant fuel cladding. Two oxide dispersion strengthened (ODS) Fe-15Cr-5Al+Y2O3 alloys were fabricated by ball milling and extrusion of gas atomized metallic powder mixed with Y2O3 powder. To assess the impact of Mo on the alloy mechanical properties, one alloy contained 1%Mo. The hardness and tensile properties of the two alloys were close and higher than the values reported for fine grain PM2000 alloy. This is likely due to the combination of a very fine grain structure and the presence of nano oxide precipitates. The nano oxide dispersion was however not sufficient to prevent grain boundary sliding at 800 C and the creep properties of the alloys were similar or only slightly superior to fine grain PM2000 alloy. Both alloys formed a protective alumina scale at 1200 C in air and steam and the mass gain curves were similar to curves generated with 12Cr-5Al+Y2O3 (+Hf or Zr) ODS alloys fabricated for a different project. To estimate the maximum temperature limit of use for the two alloys in steam, ramp tests at a rate of 5 C/min were carried out in steam. Like other ODS alloys, the two alloys showed a significant increase of the mas gains at T~ 1380 C compared with ~1480 C for wrought alloys of similar composition. The beneficial effect of Yttrium for wrought FeCrAl does not seem effective for most ODS FeCrAl alloys. Characterization of the hardness of annealed specimens revealed that the microstructure of the two alloys was not stable above 1000 C. Concurrent radiation results suggested that Cr levels <15wt% are desirable and the creep and oxidation results from the 12Cr ODS alloys indicate that a lower Cr, high strength ODS alloy with a higher maximum use temperature could be achieved.

  13. Project on New Domestic Zirconium Alloy Fuel Assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Pei-sheng; ZHANG; Ai-min

    2012-01-01

    <正>The objectives of the project is to conduct irradiation at research reactor for small fuel assembly with domestic new zirconium alloy, and then to carry out post irradiation examination, and finally to acquire

  14. Temperature measuring analysis of the nuclear reactor fuel assembly

    Science.gov (United States)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  15. Numerical Simulation for Flow Distribution in ACE7 Fuel Assemblies affected by a Spacer Grid Deformation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jongpil; Jeong, Ji Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In spite of various efforts to understand hydraulic phenomena in a rod bundle containing deformed rods due to swelling and/or ballooning of clad, the studies for flow blockage due to spacer grid deformation have been limited. In the present work, 3D CFD analysis for flow blockage was performed to evaluate coolant flow within ACE7 fuel assemblies (FAs) containing a FA affected by a spacer grid deformation. The real geometry except for inner grids was used in the simulation and the region including inner grid was replaced by porous media. In the present work, the numerical simulation was performed to predict coolant flow within ACE7 FAs affected by a Mid grid deformation. The 3D CFD result shows that approximately 60 subchannel hydraulic diameter is required to fully recover coolant flow under normal operating condition.

  16. The oxidation and hydriding of zircaloy fuel cladding in high temperature aqueous solutions

    Science.gov (United States)

    Chen, Yingzi

    Nearly 90% of today's fission reactors use Zr based fuel cladding materials. The Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) are the two most common water-cooled nuclear reactors. Corrosion is the principal threat to the failure of the fuel in these reactors, resulting in the release of fission products to the coolant and hence to the establishment of radiation fields in out-of-core regions of the coolant circuit (e.g., steam generators in PWRs and turbines in BWRs). As is well known, corrosion is an electrochemical phenomenon; however, electrochemical effects are often neglected in corrosion studies on zirconium and its alloys, because of the difficulty in performing well-defined experiments under the appropriate conditions (high temperatures and pressures). In-situ studies have been carried out to examine the electrochemistry of passive zirconium under simulated BWR and PWR coolant conditions by using a controlled hydrodynamic, high temperature/high pressure test cell. The oxidation/hydriding mechanisms are elucidated by measuring the current, impedance, and capacitance of passive zirconium as a function of formation potential. The data are interpreted in terms of a modified point defect model (PDM) that recognize the existence of a passive film comprising a thick oxide outer layer over a thin barrier layer. From the composition of the zirconium passive film and thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Transients in current density and the thickness of the passive film formed on zirconium, when stepping the potential in either the positive or negative directions, have confirmed that the rate law afforded by the PDM adequately describes the growth and thinning of the passive film at high temperatures. The experimental results demonstrate that the kinetics of either oxygen or hydrogen vacancy generation

  17. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  18. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  19. Review of qualifications for fuel assembly fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Slabu, Dan; Zemek, Martin; Hellwig, Christian [Axpo AG, Baden (Switzerland)

    2013-02-15

    The required quality of nuclear fuel in industrial production can only be assured by applying processes in fabrication and inspection, which are well mastered and have been proven by an appropriate qualification. The present contribution shows the understanding and experiences of Axpo with respect to qualifications in the frame of nuclear fuel manufacturing and reflects some related expectations of the operator. (orig.)

  20. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-02-29

    The code DIONISIO 1.0 describes most of the main phenomena occurring in a fuel rod throughout its life under normal operation conditions of a nuclear thermal reactor. Starting from the power history, DIONISIO predicts the temperature distribution in the domain, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the rod free volume, gas mixing, pressure increase, restructuring and grain growth in the UO{sub 2} pellet, irradiation growth of the Zircaloy cladding, oxide layer growth on its surface, hydrogen uptake and the effects of a corrosive atmosphere either internal or external. In particular, the models of thermal conductance of the gap and of pellet-cladding mechanical interaction incorporated to the code constitute two realistic tools. The possibility of gap closure (including partial contact between rough surfaces) and reopening during burnup is allowed. The non-linear differential equations are integrated by the finite element method in two-dimensions assuming cylindrical symmetry. Good results are obtained for the simulation of several irradiation tests.

  1. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods

    Science.gov (United States)

    Soba, Alejandro; Denis, Alicia

    2008-02-01

    The code DIONISIO 1.0 describes most of the main phenomena occurring in a fuel rod throughout its life under normal operation conditions of a nuclear thermal reactor. Starting from the power history, DIONISIO predicts the temperature distribution in the domain, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the rod free volume, gas mixing, pressure increase, restructuring and grain growth in the UO 2 pellet, irradiation growth of the Zircaloy cladding, oxide layer growth on its surface, hydrogen uptake and the effects of a corrosive atmosphere either internal or external. In particular, the models of thermal conductance of the gap and of pellet-cladding mechanical interaction incorporated to the code constitute two realistic tools. The possibility of gap closure (including partial contact between rough surfaces) and reopening during burnup is allowed. The non-linear differential equations are integrated by the finite element method in two-dimensions assuming cylindrical symmetry. Good results are obtained for the simulation of several irradiation tests.

  2. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Science.gov (United States)

    Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-01

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

  3. Hydrogen storage and integrated fuel cell assembly

    Science.gov (United States)

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  4. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  5. Improving 6061-Al Grain Growth and Penetration across HIP-Bonded Clad Interfaces in Monolithic Fuel Plates: Initial Studies

    Energy Technology Data Exchange (ETDEWEB)

    Hackenberg, Robert E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCabe, Rodney J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Edwards, Randall L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trujillo, R. Ralph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lienert, Thomas J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Forsyth, Robert T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Harada, Kiichi L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-05-06

    Grain penetration across aluminum-aluminum cladding interfaces in research reactor fuel plates is desirable and was obtained by a legacy roll-bonding process, which attained 20-80% grain penetration. Significant grain penetration in monolithic fuel plates produced by Hot Isostatic Press (HIP) fabrication processing is equally desirable but has yet to be attained. The goal of this study was to modify the 6061-Al in such a way as to promote a much greater extent of crossinterface grain penetration in monolithic fuel plates fabricated by the HIP process. This study documents the outcomes of several strategies attempted to attain this goal. The grain response was characterized using light optical microscopy (LOM) electron backscatter diffraction (EBSD) as a function of these prospective process modifications done to the aluminum prior to the HIP cycle. The strategies included (1) adding macroscopic gaps in the sandwiches to enhance Al flow, (2) adding engineering asperities to enhance Al flow, (3) adding stored energy (cold work), and (4) alternative cleaning and coating. Additionally, two aqueous cleaning methods were compared as baseline control conditions. The results of the preliminary scoping studies in all the categories are presented. In general, none of these approaches were able to obtain >10% grain penetration. Recommended future work includes further development of macroscopic grooving, transferred-arc cleaning, and combinations of these with one another and with other processes.

  6. Project Progress of New Domestic Zirconium Alloy Fuel Sub-assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Ai-min; ZHANG; Pei-sheng; LIU; Jia-zheng; LIU; Wei

    2015-01-01

    At present,the project of new domestic zirconium alloy fuel sub-assembly irradiation is ongoing according to schedule.This paper presents progress of the project such as fuel sub-assembly detailed design,manufacturing process and fuel transportation method.1 Fuel sub-assembly detailed designing

  7. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-07-11

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  8. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  9. Nuclear imaging of the fuel assembly in ignition experiments

    Energy Technology Data Exchange (ETDEWEB)

    Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; Arnold, P. A.; Ashabranner, R. C.; Atherton, L. J.; Barrios, M. A.; Batha, S.; Bell, P. M.; Benedetti, L. R.; Berger, R. L.; Bernstein, L. A.; Berzins, L. V.; Betti, R.; Bhandarkar, S. D.; Bionta, R. M.; Bleuel, D. L.; Boehly, T. R.; Bond, E. J.; Bowers, M. W.; Bradley, D. K.; Brunton, G. K.; Buckles, R. A.; Burkhart, S. C.; Burr, R. F.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Castro, C.; Celliers, P. M.; Cerjan, C. J.; Chandler, G. A.; Choate, C.; Cohen, S. J.; Collins, G. W.; Cooper, G. W.; Cox, J. R.; Cradick, J. R.; Datte, P. S.; Dewald, E. L.; Di Nicola, P.; Di Nicola, J. M.; Divol, L.; Dixit, S. N.; Dylla-Spears, R.; Dzenitis, E. G.; Eckart, M. J.; Eder, D. C.; Edgell, D. H.; Edwards, M. J.; Eggert, J. H.; Ehrlich, R. B.; Erbert, G. V.; Fair, J.; Farley, D. R.; Felker, B.; Fortner, R. J.; Frenje, J. A.; Frieders, G.; Friedrich, S.; Gatu-Johnson, M.; Gibson, C. R.; Giraldez, E.; Glebov, V. Y.; Glenn, S. M.; Glenzer, S. H.; Gururangan, G.; Haan, S. W.; Hahn, K. D.; Hammel, B. A.; Hamza, A. V.; Hartouni, E. P.; Hatarik, R.; Hatchett, S. P.; Haynam, C.; Hermann, M. R.; Herrmann, H. W.; Hicks, D. G.; Holder, J. P.; Holunga, D. M.; Horner, J. B.; Hsing, W. W.; Huang, H.; Jackson, M. C.; Jancaitis, K. S.; Kalantar, D. H.; Kauffman, R. L.; Kauffman, M. I.; Khan, S. F.; Kilkenny, J. D.; Kimbrough, J. R.; Kirkwood, R.; Kline, J. L.; Knauer, J. P.; Knittel, K. M.; Koch, J. A.; Kohut, T. R.; Kozioziemski, B. J.; Krauter, K.; Krauter, G. W.; Kritcher, A. L.; Kroll, J.; Kyrala, G. A.; Fortune, K. N. La; LaCaille, G.; Lagin, L. J.; Land, T. A.; Landen, O. L.; Larson, D. W.; Latray, D. A.; Leeper, R. J.; Lewis, T. L.; LePape, S.; Lindl, J. D.; Lowe-Webb, R. R.; Ma, T.; MacGowan, B. J.; MacKinnon, A. J.; MacPhee, A. G.; Malone, R. M.; Malsbury, T. N.; Mapoles, E.; Marshall, C. D.; Mathisen, D. G.; McKenty, P.; McNaney, J. M.; Meezan, N. B.; Michel, P.; Milovich, J. L.; Moody, J. D.; Moore, A. S.; Moran, M. J.; Moreno, K.; Moses, E. I.; Munro, D. H.; Nathan, B. R.; Nelson, A. J.; Nikroo, A.; Olson, R. E.; Orth, C.; Pak, A. E.; Palma, E. S.; Parham, T. G.; Patel, P. K.; Patterson, R. W.; Petrasso, R. D.; Prasad, R.; Ralph, J. E.; Regan, S. P.; Rinderknecht, H.; Robey, H. F.; Ross, G. F.; Ruiz, C. L.; Seguin, F. H.; Salmonson, J. D.; Sangster, T. C.; Sater, J. D.; Saunders, R. L.; Schneider, M. B.; Schneider, D. H.; Shaw, M. J.; Simanovskaia, N.; Spears, B. K.; Springer, P. T.; Stoeckl, C.; Stoeffl, W.; Suter, L. J.; Thomas, C. A.; Tommasini, R.; Town, R. P.; Traille, A. J.; Wonterghem, B. Van; Wallace, R. J.; Weaver, S.; Weber, S. V.; Wegner, P. J.; Whitman, P. K.; Widmann, K.; Widmayer, C. C.; Wood, R. D.; Young, B. K.; Zacharias, R. A.; Zylstra, A.

    2013-05-01

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models’ prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

  10. Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

    Directory of Open Access Journals (Sweden)

    Caruso Stefano

    2016-01-01

    Full Text Available The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for

  11. Reconstruction of Spent Fuel Dissolver Critical Assembly

    Institute of Scientific and Technical Information of China (English)

    LIANG; Shu-hong; ZHU; Qing-fu; ZHOU; Qi; QUAN; Yan-hui; YANG; Li-jun; LUO; Huang-da; LIU; Yang; ZHANG; Wei; ZHOU; Xiao-ping; LIU; Dong-hai

    2015-01-01

    During the twelfth Five-Year period,Reactor Physics Laboratory has taken on the research item about spent fuel dissolver critical experiment in nuclear power development project,which should be accomplished by using the uranium solution nuclear critical safety experiment device.Due to the differences of experimental content

  12. Lateral Stiffness Analysis of Fuel Assembly as Contact Condition for PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To evaluate the fuel assembly bowing in the core, the lateral stiffness analysis is needed. In the fuel assembly, there are two load pads. One is the top load pad (TLP) and the other is above the core load pad (ACLP). These load pads supply the impact surface among the fuel assemblies. In this paper, the lateral stiffness analysis of the fuel assembly as the core contact condition will be executed using the finite element method. The lateral stiffness of a fuel assembly is established by the FE method. These analysis results will be utilized in a fuel assembly bowing analysis in the core.

  13. Development of finite element analysis code SPOTBOW for prediction of local velocity and temperature fields around distorted fuel pin in LMFBR assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-05-01

    A two-dimensional steady-state distributed parameter code SPOTBOW has been developed for predicting the fine structure of cladding temperature in an liquid metal fast breeder reactor (LMFBR) fuel assembly where the deformation of fuel pins is induced by irradiation swelling, creep and thermal distortion under high burn-up operating condition. When the deformed fuel pin approaches adjacent pins and wrapper tube and comes in contact with those, the peak temperature, known as the hot spot temperature, can appear somewhere on the outer surface of the cladding. The temperature rise across the film is an important consideration in the cladding temperature analysis. Fully developed turbulent momentum and heat transfer equations based on the empirical turbulent model are solved by using the Galerkin finite element method which is suitable for the problem of the complicated boundary shape, such as the wire-wrapped fuel pin bundle. A new iteration procedure has been developed for solving the above equations by using the rise in coolant temperature, which is obtained with subchannel analysis codes, as a boundary condition. Calculated results are presented for local temperature distribution in normal and bowing pin bundle geometry, as compared with experiments. (author).

  14. Metallography and Microanalysis of Qinshan PhaseⅠ NPP Spent Fuel Rods

    Institute of Scientific and Technical Information of China (English)

    QIAN; Jin; BIAN; Wei; GUO; Li-na; GUO; Yi-fan; CHU; Feng-min; LIANG; Zheng-qiang

    2015-01-01

    Qinshan PhaseⅠNPP is a first domestic commercial PWR and its fuel rods and fuel assembly were designed and manufactured by China.In order to assess the irradiation properties of the fuel rods,8spent fuel rods which were drew out from 3fuel assemblies were transferred to CIAE hot cells for post irradiation examination(PIE)in 2014.The cladding material of the fuel

  15. Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

    Science.gov (United States)

    Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

    2017-04-01

    A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

  16. Uncertainty Analysis for OECD-NEA-UAM Benchmark Problem of TMI-1 PWR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk; Kim, S. J.; Seo, K.W.; Hwang, D. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A quantification of code uncertainty is one of main questions that is continuously asked by the regulatory body like KINS. Utility and code developers solve the issue case by case because the general answer about this question is still opened. Under the circumference, OECD-NEA has attracted the global consensus on the uncertainty quantification through the UAM benchmark program. OECD-NEA benchmark II-2 problem is a problem on the uncertainty quantification of subchannel code. It is a problem that the uncertainty of fuel temperature and ONB location on the TMI-1 fuel assembly are estimated on the transient and steady condition. In this study, the uncertainty quantification of MATRA code is performed on the problem. Workbench platform is developed to produce the large set of inputs that is needed to estimate the uncertainty quantification on the benchmark problem. Direct Monte Carlo sampling is used to the random sampling from sample PDF. Uncertainty analysis of MATRA code on OECD-NEA benchmark problem is estimated using the developed tool and MATRA code. Uncertainty analysis on OECD-NEA benchmark II-2 problem was performed to quantify the uncertainty of MATRA code. Direct Monte Carlo sampling is used to extract 2000 random parameters. Workbench program is developed to generate input files and post process of calculation results. Uncertainty affected by input parameters was estimated on the DNBR, the cladding and the coolant temperatures.

  17. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  18. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: jersilva@ipen.br; laaterre@ipen.br; myrthes@ipen.br; cteodoro@ipen.br; teixeira@ipen.br; madamy@ipen.br; glucki@ipen.br

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  19. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  20. 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)

  1. Premixer assembly for mixing air and fuel for combustion

    Energy Technology Data Exchange (ETDEWEB)

    York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

    2016-12-13

    A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

  2. Chemical compatibility between UO2 fuel and SiC cladding for LWRs. Application to ATF (Accident-Tolerant Fuels)

    Science.gov (United States)

    Braun, James; Guéneau, Christine; Alpettaz, Thierry; Sauder, Cédric; Brackx, Emmanuelle; Domenger, Renaud; Gossé, Stéphane; Balbaud-Célérier, Fanny

    2017-04-01

    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 UO2±x/SiC system is performed. Moreover, using two different experimental methods, the chemical compatibility of SiC towards uranium dioxide, with various oxygen contents (UO2±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 UO2+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.

  3. High Temperature and Pressure Steam-H2 Interaction with Candidate Advanced LWR Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL

    2012-08-01

    This report summarizes the work completed to evaluate cladding materials that could serve as improvements to Zircaloy in terms of accident tolerance. This testing involved oxidation resistance to steam or H{sub 2}-50% steam environments at 800-1350 C at 1-20 bar for short times. A selection of conventional alloys, SiC-based ceramics and model alloys were used to explore a wide range of materials options and provide guidance for future materials development work. Typically, the SiC-based ceramic materials, alumina-forming alloys and Fe-Cr alloys with {ge}25% Cr showed the best potential for oxidation resistance at {ge}1200 C. At 1350 C, FeCrAl alloys and SiC remained oxidation resistant in steam. Conventional austenitic steels do not have sufficient oxidation resistance with only {approx}18Cr-10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application and high Cr contents raise concern about {alpha}{prime} formation. Higher pressures (up to 20.7 bar) and H{sub 2} additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed in a H{sub 2}-50%H{sub 2}O environment at 10.3 bar. As some of the results regarding low-alloyed FeCrAl and Fe-Cr alloys were unexpected, further work is needed to fundamentally understand the minimum Cr and Al alloy contents needed for protective behavior in these environments in order to assist in alloy selection and guide alloy development.

  4. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    Science.gov (United States)

    Yamamoto, Y.; Pint, B. A.; Terrani, K. A.; Field, K. G.; Yang, Y.; Snead, L. L.

    2015-12-01

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10-20Cr, 3-5Al, and 0-0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741 °C.

  5. Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications

    Science.gov (United States)

    Kim, Daejong; Lee, Hyun-Geun; Park, Ji Yeon; Kim, Weon-Ju

    2015-03-01

    The SiC ceramics are under investigation for the fuel cladding in the light water nuclear reactors because of its excellent high temperature strength and corrosion resistance against hot steam under the severe accident conditions. In this study, the SiC triplex tubes consisting of a SiC inner layer, a SiC/PyC/SiC intermediate layer, and a SiC outer layer were fabricated by the chemical vapor processes. The hoop strength and fracture behaviors of the SiC triplex tube were investigated. The SiC triplex tubes fabricated at the high ratio of H2/MTS had a quite high average strength with a relatively small standard deviation. The hoop strength of the composite tubes tends to increase with the volume fraction of the reinforced fibers. The highest fiber volume fraction was obtained using Tyranno SA3-0.8k with the dense winding patterns such as bamboo-like mosaic pattern, which resulted in the high hoop strength compared to other fibers of Tyranno SA3-1.6k and Hi-Nicalon Type S. Hoop strength also increased slightly as the winding angle increased from 45° to 65°. Fracture behaviors of the SiC triplex tube were investigated via the observation of microstructure of the failed samples.

  6. Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Daejong, E-mail: dkim@kaeri.re.kr; Lee, Hyun-Geun; Park, Ji Yeon; Kim, Weon-Ju

    2015-03-15

    The SiC ceramics are under investigation for the fuel cladding in the light water nuclear reactors because of its excellent high temperature strength and corrosion resistance against hot steam under the severe accident conditions. In this study, the SiC triplex tubes consisting of a SiC inner layer, a SiC/PyC/SiC intermediate layer, and a SiC outer layer were fabricated by the chemical vapor processes. The hoop strength and fracture behaviors of the SiC triplex tube were investigated. The SiC triplex tubes fabricated at the high ratio of H{sub 2}/MTS had a quite high average strength with a relatively small standard deviation. The hoop strength of the composite tubes tends to increase with the volume fraction of the reinforced fibers. The highest fiber volume fraction was obtained using Tyranno SA3-0.8k with the dense winding patterns such as bamboo-like mosaic pattern, which resulted in the high hoop strength compared to other fibers of Tyranno SA3-1.6k and Hi-Nicalon Type S. Hoop strength also increased slightly as the winding angle increased from 45° to 65°. Fracture behaviors of the SiC triplex tube were investigated via the observation of microstructure of the failed samples.

  7. Investigations of Aluminum-Doped Self-Healing Zircaloy Surfaces in Context of Accident-Tolerant Fuel Cladding Research

    Science.gov (United States)

    Carr, James; Vasudevamurthy, Gokul; Snead, Lance; Hinderliter, Brian; Massey, Caleb

    2016-06-01

    We present here some important results investigating aluminum as an effective surface dopant for increased oxidation resistance of zircaloy nuclear fuel cladding. At first, the transport behavior of aluminum into reactor grade zircaloy was studied using simple diffusion couples at temperatures greater than 770 K. The experiments revealed the formation of tens of microns thick graded Zr-Al layers. The activation energy of aluminum in zircaloy was found to be ~175 kJ/mol (~1.8 eV), indicating the high mobility of aluminum in zircaloy. Subsequently, aluminum sputter-coated zircaloy coupons were heat-treated to achieve surface doping and form compositionally graded layers. These coupons were then tested in steam environments at 1073 and 1273 K. The microstructure of the as-fabricated and steam-corroded specimens was compared to those of pure zircaloy control specimens. Analysis of data revealed that aluminum effectively competed with zircaloy for oxygen up until 1073 K blocking oxygen penetration, with no traces of large scale spalling, indicating mechanically stable interfaces and surfaces. At the highest steam test temperatures, aluminum was observed to segregate from the Zr-Al alloy under layers and migrate to the surface forming discrete clusters. Although this is perceived as an extremely desirable phenomenon, in the current experiments, oxygen was observed to penetrate into the zirconium-rich under layers, which could be attributed to formation of surface defects such as cracks in the surface alumina layers.

  8. 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.

  9. The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Jiang, Hao [ORNL

    2016-01-12

    To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.

  10. Examination of the chemical composition of irradiated zirconium based fuel claddings at the metal/oxide interface by TEM

    Science.gov (United States)

    Abolhassani, S.; Bart, G.; Jakob, A.

    2010-04-01

    Detailed post-irradiation examinations have been performed at PSI on three fuel rods with differing cladding materials revealing different corrosion behaviour. The rods had been irradiated for 3-5 cycles at Gösgen nuclear power plant (pressurised water reactor), Switzerland. As zirconium corrosion is proceeding at the metal/oxide interface, extended micro-structural analyses were performed by transmission electron microscopy (TEM), expecting to possibly reveal phenomena explaining the varying corrosion resistance. This paper reports on the distribution of oxygen at the metal/oxide interface examined by energy dispersive X-ray spectroscopy (EDS) in TEM, while other micro-structural investigations have been published earlier [1]. In order to get some statistical confidence in the analyses, three neighbouring TEM samples of each cladding variant were studied. The oxygen concentration profiles of the three alloys (i.e. low-tin Zircaloy-4, Zr2.5%Nb and extra low-tin (Sn 0.56%)) both in the oxide and metal close to the metal/oxide interface are compared. The results of the examinations show the composition of the oxide in the vicinity of the interface to be sub-stoichiometric for all three materials, indicating an oxide layer adjacent to the interface, with diffusion-controlled access of oxygen to the metal/oxide interface. The metallic parts show highest oxygen concentrations at the metal/oxide interface which are reduced towards the bulk metal, pointing towards the expected second diffusion-controlled process leading to α-Zr (O). Based on the experimental results values for the diffusion coefficients in the range of 0.8-6.0 × 10 -20 m 2 s -1 are estimated for the oxygen dissolution process, the diffusion coefficient in Zircaloy-4 being six times higher than for the other two less corroding alloys. This finding is in contradiction with the present assumptions about the corrosion mechanism, and confirms the expected but not so far reported diffusion controlled

  11. Protective Coatings for Wet Storage of Aluminium-Clad Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, S.M.C.; Correa, O.V.; Souza, J.A. De; Ramanathan, L.V. [Materials science and Technology Center, Instituto de Pesquisas Energeticas e Nucleares - IPEN, Av. Prof. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

    2011-07-01

    Corrosion protection of spent RR fuel for long term wet storage was considered important, primarily from the safety standpoint and the use of conversion coatings was proposed in 2008. This paper presents the results of: (a) on-going field tests in which un-coated and lanthanide-based conversion coated Al alloy coupons were exposed to the IEA-R1 reactor spent fuel basin for durations of up to a year; (b) preparation of cerium modified hydrotalcite coatings and cerium sealed boehmite coatings on AA 6061 alloy; (c) corrosion resistance of coated specimens in NaCl solutions. The field studies indicated that the oxidized and cerium dioxide coated coupons were the most corrosion resistant. The cerium modified hydrotalcite and cerium sealed boehmite coated specimens showed marked increase in pitting corrosion resistance. (author)

  12. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    Directory of Open Access Journals (Sweden)

    Bobrov Evgenii

    2016-01-01

    Full Text Available This paper shows basic features of different fuel assembly (FA application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling.

  13. PWR-2 Blanket Fuel Assembly Removal Safety Basis Criteria Document

    Energy Technology Data Exchange (ETDEWEB)

    BUSHORE, R.P.

    2001-01-22

    This criteria document describes the proposed format, content, and schedule for the preparation of an amendment to the Interim Safety Basis for Solid Waste Facilities (T Plant) (ISB), (HNF-SD-WM-ISB-006), and to the T Plant Interim Operational Safety Requirements (IOSR) (''F-SD-WM-TSR-003). The amendments to these documents are intended to authorize removal of spent nuclear fuel (SNF) assemblies from the spent fuel pool in the Solid Waste Treatment Facility 221-T canyon for interim storage in the Canister Storage Building (CSB). The amendments will include a stand-alone safety assessment as well as revisions to these safety documents as needed to reflect the changes in work scope not currently authorized to accomplish the expected end-state of the Fuel Removal Project for the 221-T Facility.

  14. Robustness in the design and manufacture of the AP1000 fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sumit, Ray [New Reactor Fuel Engineering, Westinghouse Electric Corporation, Monroeville, PA (United States)

    2009-06-15

    corrosion risk assessments have been performed for the AP1000 cores to demonstrate that these are within the experience base of existing designs. These results will be presented in the paper. In addition, specific additional actions, including Chemistry specifications, and Zinc addition, will be discussed. 5) Manufacturing: A summary of manufacturing process improvements that are being implemented on the AP1000 production campaign will be discussed. Specific focus will be placed on improvements to pellet and clad quality: however, improvements to other aspects of the fuel assembly will also be discussed. Finally, the paper will discuss the rationale behind the planning process that is underway for the performance of fuel surveillance and inspection of the initial AP1000 cores. (authors)

  15. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  16. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, A.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

  17. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  18. Rail Shock and Vibration Pre-Test Modeling of a Used Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B.; Klymyshyn, Nicholas A.; Jensen, Philip J.; Best, Ralph E.; Maheras, Steven J.; McConnell, Paul E.; Orchard, John

    2015-04-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Available information is not sufficient to determine the ability of ES UNF, including high-burnup fuel, to withstand shock and vibration forces that could occur when the UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility. There are three major gaps in the available information – 1) the forces that UNF assemblies would be subjected to when transported by rail, 2) the mechanical characteristics of fuel rod cladding, which is an essential structure for controlling the geometry of the UNF, a safety related feature, and 3) modeling methodologies to evaluate multiple possible degradation or damage mechanisms over the UNF lifetime. In order to address the first gap, options for tests to determine the physical response of surrogate UNF assemblies subjected to shock and vibration forces that are expected to be experienced during normal conditions of transportation (NCT) by rail must be identified and evaluated. The objective of the rail shock and vibration tests is to obtain data that will help researchers understand the mechanical loads that ES UNF assemblies would be subjected to under normal conditions of transportation and to fortify the computer modeling that will be necessary to evaluate the impact

  19. Fuel assembly design for APR1400 with low CBC

    Energy Technology Data Exchange (ETDEWEB)

    Hah, Chang Joo, E-mail: changhah@kings.ac.kr [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  20. Fuel assembly design for APR1400 with low CBC

    Science.gov (United States)

    Hah, Chang Joo

    2015-04-01

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to ΔkTARGET. A set of new designed fuel assembly satisfies an objective function, min [f =∑i (ΔkF A-Δki ) ] , and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to ΔkTARGET as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  1. Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    2015-12-01

    As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

  2. Out-of-pile Verifying Test for the Hydraulic Stability of the CARR Standard Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The CARR standard fuel element is a flat-plate-type assembly. A fuel plate consists of 0.6 mmthickness layer of uranium- silicon - aluminum fuel (U3Si2-Al) and 0.38 mm thickness of aluminumcladding. The fuel plates are attached to aluminum alloy side plates by a "roll swaging" technique. Thistype of fuel assembly is first used in China. The testing simulates the in-pile thermal-hydraulic operating conditions except for neutron

  3. Evaluation of aluminum-clad spent fuel corrosion in Argentine basins

    Energy Technology Data Exchange (ETDEWEB)

    Haddad, R.; Loberse, A.N.; Semino, C.J.; Guasp, R. [CNEA, Buenos Aires (Argentina)

    2001-07-01

    An IAEA sponsored Coordinated Research Program was extended to study corrosion effects in several sites. Racks containing Aluminum samples were placed in different positions of each basin and periodic sampling of all the waters was performed to conduct chemical analysis. Different forms of corrosion have been encountered during the programme. In general, the degree of degradation is inversely proportional to the purity of the water. Maximum pit depths after 2 years of exposure are in the range of 100-200 {mu}m. However, sediments deposited on the coupon surfaces seem to be responsible for the developing of large pits (1-2 mm in diameter). In many cases, what appears to be iron oxide particles were found originated by the corrosion of carbon steel components present elsewhere in the basin. These results correlate with observations made on the fuel itself, during exhaustive visual inspection. (author)

  4. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  5. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  6. Results of Severe Fuel Damage Experiment QUENCH-14 with Advanced Rod Cladding M5®. (KIT Scientific Reports ; 7549)

    OpenAIRE

    STUCKERT J.; Große, M.; Stegmaier, U.; Steinbrück, M.

    2010-01-01

    The QUENCH experiments are to investigate the hydrogen release resulting from the water injection into an uncovered core of a Light Water Reactor as well as the high-temperature behavior of core materials. The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with the tests QUENCH-06 that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings.

  7. Development of Tools for Treating an Irradiated Fuel Rod Assembly in the Pool of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. T.; Ahn, S. H.; Kim, K. H.; Joung, C. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    To inspect a fuel rod during irradiation testing at the test loop of a research reactor, the test rig should be disassembled from the IPS (In-pile test section), and the targeted fuel rod assembly should be disassembled from the test rig and encapsulated in a cask to deliver the assembly to the hot cell. In addition, the fuel rod assembly under inspection in the hot cell should be delivered to the reactor pool and reassembled into the test rig to resume the irradiation test. Because the irradiated fuel rod is highly radioactive, all of the assembly and disassembly operations should be carried out in the reactor pool. Therefore, special tools need to be developed to treat the test rig in the pool of a research reactor. In this study, a new mechanically detachable fuel rod assembly has been developed for intermediate inspection during irradiation test at HANARO. A fuel rod assembly can be divided into two parts, such as an instrumented fuel rod assembly and a non-instrumented fuel rod assembly. In particular, an instrumented fuel rod assembly is assembled at the lower part of the test rig, and a non-instrumented fuel rod assembly is assembled at the bottom of the instrumented fuel rod assembly. The non-instrumented fuel rod assembly is locked in the test rig during irradiation test, and is easily disassembled from the instrumented fuel rod assembly by pushing the anchor button and twisting the non-instrumented fuel rod assembly. In addition, because a test rig is 5.4 meters long and the disassembling operation should be carried out at 6 meters deep in the pool of HANARO, tools to help disassemble and assemble the non-instrumented fuel rod assembly have also been developed. All components were designed to operate mechanically and are made of stainless steel and Al 6061 to minimize the effects from the radioactivity. The performance of the developed fuel rod assembly and tools have been verified through an out pile test.

  8. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  9. Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding

    Science.gov (United States)

    Alat, Ece; Motta, Arthur T.; Comstock, Robert J.; Partezana, Jonna M.; Wolfe, Douglas E.

    2016-09-01

    In an attempt to develop an accident-tolerant fuel (ATF) that can delay the deleterious consequences of loss-of-coolant-accidents (LOCA), multilayer coatings were deposited onto ZIRLO® coupon substrates by cathodic arc physical vapor deposition (CA-PVD). Coatings were composed of alternating TiN (top) and Ti1-xAlxN (2-layer, 4-layer, 8-layer and 16-layer) layers. The minimum TiN top coating thickness and coating architecture were optimized for good corrosion and oxidation resistance. Corrosion tests were performed in static pure water at 360 °C and 18.7 MPa for up to 90 days. The optimized coatings showed no spallation/delamination and had a maximum of 6 mg/dm2 weight gain, which is 6 times smaller than that of a control sample of uncoated ZIRLO® which showed a weight gain of 40.2 mg/dm2. The optimized architecture features a ∼1 μm TiN top layer to prevent boehmite phase formation during corrosion and a TiN/TiAlN 8-layer architecture which provides the best corrosion performance.

  10. Development of laser welded appendages to Zircaloy-4 fuel tubing (sheath/cladding)

    Energy Technology Data Exchange (ETDEWEB)

    Livingstone, S., E-mail: steve.livingstone@cnl.ca [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Xiao, L. [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Corcoran, E.C.; Ferrier, G.A.; Potter, K.N. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, ON, Canada K7K 7B4 (Canada)

    2015-04-01

    Highlights: • Examines feasibility of laser welding appendages to Zr-4 tubing. • Laser welding minimizes the HAZ and removes toxic Be. • Mechanical properties of laser welds appear competitive with induction brazed joints. • Work appears promising and lays the foundation for further investigations. - Abstract: Laser welding is a potential alternative to the induction brazing process commonly used for appendage attachment in CANDU{sup ®} fuel fabrication that uses toxic Be as a filler metal, and creates multiple large heat affected zones in the sheath. For this work, several appendages were laser welded to tubing using different laser heat input settings and then examined with a variety of techniques: visual examination, metallography, shear strength testing, impact testing, and fracture surface analysis. Where possible, the examination results are contrasted against production induction brazed joints. The work to date looks promising for laser welded appendages. Further work on joint optimization, corrosion testing, irradiation testing, and post-irradiation examination will be performed in the future.

  11. Ultrasonic Bonding of Membrane-Electrode-Assemblies of Fuel Cells

    Directory of Open Access Journals (Sweden)

    Dung-An Wang

    2016-05-01

    Full Text Available Ultrasonic bonding has a great potential for manufacturing of membrane electrode assemblies (MEAs of fuel cells (FCs due to its short process cycle time and low energy consumption.  Before introduction of the bonding process into the industry, a detailed and elaborate investigation of the effects of the processing parameters on the bonding quality is necessary.  We develop a finite element model of the ultrasonic bonding for MEAs of FCs.  The model can be used as a computational framework for initial evaluation of the effectiveness of ultrasonic boding for MEAs of FCs.

  12. Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Alat, Ece, E-mail: exa179@psu.edu [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Motta, Arthur T. [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Comstock, Robert J.; Partezana, Jonna M. [Westinghouse Electric Co., Beulah Rd, Pittsburgh, PA 1332 (United States); Wolfe, Douglas E. [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Applied Research Laboratory, The Pennsylvania State University, 119 Materials Research Building, University Park, PA 16802 (United States)

    2016-09-15

    In an attempt to develop an accident-tolerant fuel (ATF) that can delay the deleterious consequences of loss-of-coolant-accidents (LOCA), multilayer coatings were deposited onto ZIRLO{sup ®} coupon substrates by cathodic arc physical vapor deposition (CA-PVD). Coatings were composed of alternating TiN (top) and Ti{sub 1-x}Al{sub x}N (2-layer, 4-layer, 8-layer and 16-layer) layers. The minimum TiN top coating thickness and coating architecture were optimized for good corrosion and oxidation resistance. Corrosion tests were performed in static pure water at 360 °C and 18.7 MPa for up to 90 days. The optimized coatings showed no spallation/delamination and had a maximum of 6 mg/dm{sup 2} weight gain, which is 6 times smaller than that of a control sample of uncoated ZIRLO{sup ®} which showed a weight gain of 40.2 mg/dm{sup 2}. The optimized architecture features a ∼1 μm TiN top layer to prevent boehmite phase formation during corrosion and a TiN/TiAlN 8-layer architecture which provides the best corrosion performance. - Highlights: • The first study on multilayer TiAlN and TiN ceramic coatings on ZIRLO{sup ®} coupons. • Corrosion tests were performed at 360°C and 18.7 MPa for up to 90 days. • Coatings adhered well to the substrate, and showed no spallation/delamination. • Weight gains were six times lower than those of uncoated ZIRLO{sup ®} samples. • Longer and higher temperature corrosion tests will be discussed in a further paper.

  13. Modelling cladding response to changing conditions

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  14. Effects of heat transfer coefficient treatments on thermal shock fracture prediction for LWR fuel claddings in water quenching

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho; Lee, Jeong Ik; Cheon, Hee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Accurate modeling of thermal shock induced stresses has become ever most important to emerging accident-tolerant ceramic cladding concepts, such as silicon carbide (SiC) and SiC coated zircaloy. Since fractures of ceramic (entirely ceramic or coated) occur by excessive tensile stresses with linear elasticity, modeling transient stress distribution in the material provides a direct indication of the structural integrity. Indeed, even for the current zircaloy cladding material, the oxide layer formed on the surface - where cracks starts to develop upon water quenching - essentially behaves as a brittle ceramic. Hence, enhanced understanding of thermal shock fracture of a brittle material would fundamentally contribute to safety of nuclear reactors for both the current fuel design and that of the coming future. Understanding thermal shock fracture of a brittle material requires heat transfer rate between the solid and the fluid for transient temperature fields of the solid, and structural response of the solid under the obtained transient temperature fields. In water quenching, a solid experiences dynamic time-varying heat transfer rates with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates during the water quenching transience has been overlooked in assessments of mechanisms, predictability, and uncertainties for thermal shock fracture. Rather, a time-constant heat transfer coefficient, named 'effective heat transfer coefficient' has become a conventional input to thermal shock fracture analysis. No single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic heat transfer coefficient changes with fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials and complete the picture of stress evolution in the quenched solid. The presented result

  15. Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tome, Carlos N [Los Alamos National Laboratory; Caro, J A [Los Alamos National Laboratory; Lebensohn, R A [Los Alamos National Laboratory; Unal, Cetin [Los Alamos National Laboratory; Arsenlis, A [LLNL; Marian, J [LLNL; Pasamehmetoglu, K [INL

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

  16. Evaluation of the effect of B and N on the microstructure of 9Cr-2W steel during an aging treatment for SFR fuel cladding tubes

    Science.gov (United States)

    Jeong, Eun Hee; Park, Sang-Gyu; Kim, Sung Ho; Kim, Young Do

    2015-12-01

    In this study, the microstructure of sodium-cooled fast reactor (SFR) fuel cladding steel with different B and N contents after aging is compared. The addition of nitrogen produces a large quantity of MX precipitates with sizes of 0.1 μm or smaller during the initial thermal treatment process and this contributes to help such precipitates maintain stability without being excessively affected by aging. B is primarily distributed in the grain boundary precipitates and grain interior precipitates in the initial stage. The B distribution is believed to move to the Cr precipitates after 7000 h and to contribute to suppressing the growth of M23C6.

  17. Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding

    Science.gov (United States)

    Stone, J. G.; Schleicher, R.; Deck, C. P.; Jacobsen, G. M.; Khalifa, H. E.; Back, C. A.

    2015-11-01

    Silicon carbide (SiC) fiber, SiC matrix composites (SiC/SiC) are being considered as a cladding material for light water reactors in order to improve safety performance. Engineered, multi-layer cladding designs consisting of both monolithic SiC (mSiC) and SiC/SiC have been examined as promising concepts to meet both strength and impermeability requirements. A new model has been developed to calculate stresses and failure probabilities for multi-layer cladding consisting of SiC-based materials in reactor operating conditions. The results show that stresses in SiC-based cladding are dominated by temperature-dependent irradiation-induced swelling, with the largest stresses occurring during the cold shutdown conditions. Failure probabilities are driven by the resulting tensile stresses at the cladding inner wall, while the outer wall is subject to compressive stresses. This indicates that the inner SiC/SiC, outer mSiC concept has the lowest failure probability, as the pseudo-plastic deformation of the composite reduces tensile loading and the compressed monolith provides a reliable, impermeable barrier to fission product release.

  18. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shore, Lawrence (Edison, NJ); Matlin, Ramail (Berkeley Heights, NJ)

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  19. The Role of X-Ray Diffraction for Analyzing Zr-Sn-Nb-Fe Alloys as Power Reactor Fuel Cladding

    Directory of Open Access Journals (Sweden)

    Sugondo

    2010-08-01

    Full Text Available Synthesis of Zr-1%Nb-1%Sn-1%Fe alloy is undertaken in order to develop fuel cladding alloy at high burn-up. Powder specimens of Zr-Sn-Nb-Fe alloy were prepared and then formed into pellets with a dimension of 10 mm in height 10 mm in diameter using a pressure of 1.2 ton/cm2. The 5 gram green pellets were then melted in an arc furnace crucible under argon atmosphere. The pressure in the furnace was set at 2 psi and the current was 50 A. Afterwards, the ingots were heated at a temperature of 1100°C for 2 hours and subsequently quenched in water. The ingots then underwent annealing at temperatures of 400°C, 500°C, 600°C, 700°C, and 750°C for 2 hours. The specimens were analyzed using X-ray diffraction in order to construct diffractograms. Results of the diffraction patterns were fitted with data from JCPDF (Joint Committee Powder Diffraction File to determine the type of crystals in the elements or substances. The greater the crystallite dimension, the smaller the dislocation density. Agreeable results for hardening or strengthening were obtained at annealing temperatures of 500°C and 700, whereas for softening or residual stress at 600°C and 750°C. The nucleation of the secondary phase precipitate (SPP was favourable at annealing temperatures of 400°C, 500°C, and 700°C. For Zr-1%Nb-1%Sn-1%Fe alloy with annealing temperatures between 400°C to 800°C, precipitates of Fe2Nb, ZrSn2,FeSn, SnZr, NbSn2, Zr0.68Nb0.25Fe0.08, Fe2Nb0.4Zr0.6, Fe37Nb9Zr54, and ω-Zr were observed. Satisfactory precipitate stabilization was achieved at annealing temperature of 800°C, growth of precipitates at temperature between 500°C to 600°C, and minimization of precipitate size at 700°C.

  20. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Directory of Open Access Journals (Sweden)

    Panferov Pavel

    2016-01-01

    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  1. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Science.gov (United States)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  2. Effect of Heat treatment and Aging Conditions on the Microstructure and Mechanical Properties of HT9 Steel for Fuel Cladding Tube

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyeong Min; Kim, Jong Lyeol [Hanyang university, Ansan (Korea, Republic of); Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel cladding tube is the most important safety barrier in a fission nuclear reactor. Thermal creep and void swelling occur by fission gas at high temperature during service time. Ferritic-martensitic steels are being considered attractive candidate materials for a fuel cladding of the SFR owing to their low expansion coefficients, high thermal conductivities and excellent irradiation resistance to void swelling. However, HT9 steel has a problem of a relatively low high temperature strength and low creep strength. To solve this problem, a study was conducted to increase the high temperature strength by changing the intermediate heat treatment step in the fabrication process of ferritic martensitic steel and controlling the microstructure and precipitate within the material. 700-780 .deg. C contributed to the increase in precipitate size, and the decrease in yield stress and hardness. An empirical equation for the mechanical properties of HT9 was suggested as a function of the microstructure and Hollomon-Jaffe tempering parameter. The results show that the size of the carbide and lath increased after aging, whereas the size of the prior austenite grain was not changed. Both the strength and hardness were decreased with aging, and this tendency saturated after 3000 hours of aging.

  3. 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.

  4. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Science.gov (United States)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  5. Cap assembly for a bundled tube fuel injector

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  6. High Energy X-ray Study on Nondestructive Detection of Fuel Assemblies

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Xiang-yang; WANG; Guo-bao; HE; Gao-kui; CUI; Yao; ZENG; Zi-qiang; ZHANG; Li-feng; LIANG; Zheng-qiang; YIN; Zhen-guo; WANG; Xin

    2015-01-01

    Nuclear fuel assemblies are the core of nuclear facilities,and the safety and effective operation of nuclear fuel assembly under complicated environment in reactor is the most important issue of guarantee of nuclear facility.In order to better research and analyze complete behavior of nuclear

  7. A spray cooling technique for spent fuel assembly stored in pool

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Dao-Gang; Cao, Q. [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Wang, Y.; Zhong, Hao-Liang; Duan, Xiao-Han

    2016-05-15

    For the safety of spent nuclear fuel assemblies stored in storage pool in the extreme condition where the water is lost completely, a passive spray cooling technique was designed, and its effectiveness has been validated by a functional experiment. The spray cooling characteristics of the spent fuel assembly have also been investigated by the experiment.

  8. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Energy Technology Data Exchange (ETDEWEB)

    Viererbl, L., E-mail: vie@ujv.cz [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Research Centre Rez Ltd. (Czech Republic); Lahodova, Z.; Voljanskij, A.; Klupak, V.; Koleska, M. [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Research Centre Rez Ltd. (Czech Republic); Cabalka, M. [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Turek, K. [Nuclear Physics Institute, Academy of Sciences of the Czech Republic (Czech Republic)

    2011-10-01

    During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of {sup 235}U, {sup 238}U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

  9. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Science.gov (United States)

    Viererbl, L.; Lahodová, Z.; Voljanskij, A.; Klupák, V.; Koleška, M.; Cabalka, M.; Turek, K.

    2011-10-01

    During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of 235U, 238U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

  10. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-12-31

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating moderator downward through channels in cylindrical fuel tubes. Powers were limited to prevent a flow excursion from occurring in one or more of these parallel channels. During full-power operation, limits prevented a boiling flow excursion from taking place. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increases beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of historical levels.

  11. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-09-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  12. Influence of Spacer Grid Outer Strap on Fuel Assembly Thermal Hydraulic Performance

    Directory of Open Access Journals (Sweden)

    Jingwen Yan

    2014-01-01

    Full Text Available The outer strap as a typical structure of a spacer grid enhances the mechanical strength, decreases hang-up susceptibility, and also influences thermal hydraulic performance, for example, pressure loss, mixing performance, and flow distribution. In the present study, a typical grid spacer with different outer strap designs is adopted to investigate the influence of outer strap design on fuel assembly thermal hydraulic performance by using a commercial computational fluid dynamics (CFD code, ANSYS CFX, and a subchannel analysis code, FLICA. To simulate the outer straps’ influence between fuel assemblies downstream, four quarter-bundles from neighboring fuel assemblies are constructed to form the computational domain. The results show that the outer strap design has a major impact on cross-flow between fuel assemblies and temperature distribution within the fuel assembly.

  13. The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

    Science.gov (United States)

    Rodionov, Yu. A.; Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.

    2014-03-01

    On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of 58Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.

  14. 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

  15. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  16. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod; Contribution a la modelisation du comportement mecanique des combustibles REP sous irradiation, avec en particulier le traitement de l`interaction pastille-gaine dans un crayon combustible

    Energy Technology Data Exchange (ETDEWEB)

    Hourdequin, N.

    1995-05-01

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR`s operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends.

  17. Design improvement for fretting-wear reduction of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R

    2000-06-01

    In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

  18. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Science.gov (United States)

    Grosse, M.; Steinbrueck, M.; Kaestner, A.

    2011-09-01

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for β-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  19. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M., E-mail: Mirco.Grosse@KIT.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Steinbrueck, M. [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Kaestner, A. [Department of Spallation Source, Paul Scherrer Institute (PSI), CH-5232 Villigen (Switzerland)

    2011-09-21

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for {beta}-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  20. Preliminary Design of U-Mo Alloy Dispersion Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>As a kind of new type fuel for research reactor, high density U-Mo alloy dispersion fuel which will substitute current fuel in the future is being studied and developed by RERTR. There are two characteristics

  1. Estimation of the nuclear fuel assembly eigenfrequencies in the probability sense

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2014-12-01

    Full Text Available The paper deals with upper and lower limits estimation of the nuclear fuel assembly eigenfrequencies, whose design and operation parameters are random variables. Each parameter is defined by its mean value and standard deviation or by a range of values. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of fuel assembly eigenfrequencies in the probability sense. Presented analytical approach used for the calculation of eigenfrequencies sensitivity is based on the modal synthesis method and the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and load-bearing skeleton linked by spacer grids. The method is applied for the Russian TVSA-T fuel assembly in the WWER1000/320 type reactor core in the Czech nuclear power plant Temelín.

  2. Fuel nozzle assembly for use as structural support for a duct structure in a combustor of a gas turbine engine

    Science.gov (United States)

    Wiebe, David J; Fox, Timothy A

    2015-03-31

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passage is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.

  3. Evaluation of the effect of B and N on the microstructure of 9Cr–2W steel during an aging treatment for SFR fuel cladding tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Eun Hee [Hanyang University, Department Materials Science and Engineering, 222, Wangsimni-ro, Seongdong-gu, Seoul, 133-791 (Korea, Republic of); Park, Sang-Gyu; Kim, Sung Ho [KAERI, Advanced Fuel Development Division, 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Young Do, E-mail: ydkim1@hanyang.ac.kr [Hanyang University, Department Materials Science and Engineering, 222, Wangsimni-ro, Seongdong-gu, Seoul, 133-791 (Korea, Republic of)

    2015-12-15

    In this study, the microstructure of sodium-cooled fast reactor (SFR) fuel cladding steel with different B and N contents after aging is compared. The addition of nitrogen produces a large quantity of MX precipitates with sizes of 0.1 μm or smaller during the initial thermal treatment process and this contributes to help such precipitates maintain stability without being excessively affected by aging. B is primarily distributed in the grain boundary precipitates and grain interior precipitates in the initial stage. The B distribution is believed to move to the Cr precipitates after 7000 h and to contribute to suppressing the growth of M{sub 23}C{sub 6}.

  4. Thermal analysis of IRT-T reactor fuel elements

    OpenAIRE

    Naymushin, Artem Georgievich; Chertkov, Yuri Borisovich; Lebedev, Ivan Igorevich; Anikin, Mikhail Nikolaevich

    2015-01-01

    The article describes the method and results of thermo-physical calculations of IRT-T reactor core. Heat fluxes, temperatures of cladding, fuel meat and coolant were calculated for height of core, azimuth directions of FA and each fuel elements in FA. Average calculated values of uniformity factor of energy release distribution for height of fuel assemblies were shown in this research. Onset nucleate boiling temperature and ONB-ratio were calculated. Shows that temperature regimes of fuel ele...

  5. Development of numerical models for Monte Carlo simulations of Th-Pb fuel assembly

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2017-01-01

    Full Text Available The thorium-uranium fuel cycle is a promising alternative against uranium-plutonium fuel cycle, but it demands many advanced research before starting its industrial application in commercial nuclear reactors. The paper presents the development of the thorium-lead (Th-Pb fuel assembly numerical models for the integral irradiation experiments. The Th-Pb assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design of the assembly allows different combinations of rods for various types of irradiations and experimental measurements. The numerical model of the Th-Pb assembly was designed for the numerical simulations with the continuous energy Monte Carlo Burnup code (MCB implemented on the supercomputer Prometheus of the Academic Computer Centre Cyfronet AGH.

  6. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    Energy Technology Data Exchange (ETDEWEB)

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  7. 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)

  8. Development status and research directions on the structural components of the fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Nam; Jeong, Yeon Ho; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Bang, Jae Keon

    1997-06-01

    Survey on the structural components of the state-of-the art of the PWR fuel assembly developed by various nuclear fuel vendors has been performed. As a result, some developmental directions and mechanical/structural basic technology to be established for these structural components have been drawn out. The developmental directions are as follows; The top end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function of debris protection. The spacer grid shall be designed in shape to have a function of enhancing the thermal margin and maintaining the fuel rod integrity without fuel failure due to fuel rod fretting and vibration. The mechanical/structural basic technology which must be established is as follows; The stress analysis results shall comply with the stress criteria specified in the ASME code stress limits and the shape optimization technology shall be developed for the top/bottom end pieces. For the spacer grid cell, the nonlinear analysis model of the fuel rod and the analysis model on the flow-induced fuel rod vibration, and a study of the mechanism and a quantified model on the fuel rod fretting wear shall be developed. In addition, numerical analysis model to estimate the buckling strength of the spacer grid assembly shall be developed. Besides above technology, technology related the verification test should be developed. (author). 30 figs., 54 refs.

  9. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-04-21

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  10. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

    Full Text Available Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  11. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dongseok; Shin, Changhwan; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others.

  12. Neutron Imaging Investigations of the Secondary Hydriding of Nuclear Fuel Cladding Alloys during Loss of Coolant Accidents

    Science.gov (United States)

    Grosse, M.; Roessger, C.; Stuckert, J.; Steinbrueck, M.; Kaestner, A.; Kardjilov, N.; Schillinger, B.

    The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrogen and zirconium was determined by testing calibration specimens with known hydrogen concentrations. Hydrogen enrichments located at the end of the ballooning zone of the tested tubes were detected in the inner rods of the test bundles. Nearly all of the peripheral claddings exposed to lower temperatures do not show such enrichments. This implies that under the conditions investigated a threshold temperature exists below which no hydrogen enrichments can be formed. In order to understand the hydrogen distribution a model was developed describing the processes occurring during loss of coolant accidents after rod burst. The general shape of the hydrogen distributions with a peak each side of the ballooning region is well predicted by this model whereas the absolute concentrations are underestimated compared to the results of the neutron tomography investigations. The model was also used to discuss the influence of the alloy composition on the secondary hydrogenation. Whereas the relations for the maximal hydrogen concentrations agree well for one and the same alloy, the agreement for tests with different alloys is less satisfying, showing that material parameters such as oxidation kinetics, phase transition temperature for the zirconium oxide, and yield strength and ductility at high temperature have to be taken into account to reproduce the results of neutron imaging investigations correctly.

  13. 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.

  14. Chemical thermodynamics of the system Cs--U--Zr--H--I--O in the light water reactor fuel-cladding gap

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, T.M.; Lindemer, T.B.

    1978-10-01

    Equilibrium thermodynamic calculations were performed on the Cs-U-Zr-H-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor fuel under in-reactor, steam, and 50% steam--50% air conditions. The in-reactor oxygen potential is assumed to be controlled by either UO/sub 2+x/ + Cs/sub 2/UO/sub 4/ or Zr + ZrO/sub 2/. The important condensed phases in-reactor are UO/sub 2+x/, Cs/sub 2/UO/sub 4/, and CsI, and the major gaseous species are Cs, Cs/sub 2/, CsI, and Cs/sub 2/I/sub 2/. The presence of steam does not alter these species, although CsOH also becomes a major gaseous species. In a 50% steam--50% air mixture, the equilibrium condensed phases are U/sub 3/O/sub 8/ or UO/sub 3/ and Cs/sub 2/U/sub 15/O/sub 46/. Under a nonequilibrium situation where zirconium metal can react with iodine, ZrO/sub 3/ or liquid ZrI/sub 2/ is present, and the gaseous species ZrI/sub 3/ and ZrI/sub 4/ have large partial pressures.

  15. Physics Design of Criticality Assembly in Experimental Research About Criticality Safety in Spent Fuel Dissolver

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Qi

    2012-01-01

    <正>In order to meet the experimental demand of criticality safety research in the spent fuel dissolver, we need to design a suitable criticality assembly. The key problem of the design work is the core design because there are many limits for it such as the number of fuel rods loaded, fissile materials existed in the solution, reactivity control, core size and etc.

  16. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  17. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    Energy Technology Data Exchange (ETDEWEB)

    Mertyurek, Ugur, E-mail: mertyureku@ornl.gov; Gauld, Ian C., E-mail: gauldi@ornl.gov

    2016-02-15

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  18. Characterization of PEM fuel cell membrane-electrode-assemblies by electrochemical methods and microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Borup, R.L.; Vanderborgh, N.E.

    1995-05-01

    Hydrogen adsorption/desorption and CO oxidation are used to evaluate the active Pt surface area of fuel cell membrane electrode assemblies. The membrane electrode assemblies are evaluated for useful catalyst life and are examined for relative CO and CO{sub 2} tolerance. The electrochemical measurements combined with microanalysis of membrane electrode assemblies, including SEM and EDS allow a greater understanding and optimization of process variables.

  19. 非对称热轧高品质不锈钢复合板可行性模拟研究%Simulation research on feasibility of hot rolling asymmetrically assembled high quality stainless clad steel plates

    Institute of Scientific and Technical Information of China (English)

    张心金; 何冰冷; 祝志超; 何毅; 李萌蘖

    2016-01-01

    In this paper,an extra-thick clad steel plate and a thin clad steel plate were asymmetrically assembled,and the strains,contact stresses and temperature distributions of assembly during the hot rolling process were calculated by ABAQUS finite software.Through temperature compensation and cooling control,the feasibility of hot rolling asymmetrically assembled clad steel plates was also inves-tigated by numerical simulation.The results show that asymmetrical assembly design is beneficial to the interfacial bonding between carbon steel layer and stainless steel layer of the extra-thick clad plate during each rolling pass.By controlling the temperature difference between upper and lower surface, the slab warping phenomenon can be effectively improved,and both a thin and an-extra thick stainless clad steel plates can be obtained at the same time,thus improving the production efficiency.Moreo-ver,asymmetrical assembly design also enables the control of rolling and cooling,and therefore en-sures the cooperative deformation of stainless steel and carbon steel in the core section and promotes its interfical bonding as a consequence.%利用特厚规格复合板与较薄规格复合板进行非对称组坯,采用 ABAQUS 有限元软件对其热轧过程中的应变、接触应力及温度分布进行计算,并通过温度补偿及冷却控制的手段,对热轧非对称复合坯的可行性进行模拟分析。结果表明,采用非对称组坯设计,有利于特厚复合板碳钢层与不锈钢层在各道次轧制中的界面结合;通过控制复合坯上、下表面的温差,能有效改善板坯翘曲现象,并可一次性获得一块宽幅特厚复合板与一块宽幅较薄规格复合板,提高生产效率;此外,采用非对称组坯设计还可实现控轧控冷,保证芯部不锈钢与碳钢的协同变形,促进其界面结合。

  20. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  1. Verification of 235U enrichment of fresh VVER-440 fuel assemblies.

    Science.gov (United States)

    Almási, I; Nguyen, C T; Zsigrai, J; Lakosi, L; Hlavathy, Z; Nagy, P; Buglyó, N

    2012-10-01

    Enrichment of uniformly and non-uniformly enriched ("profiled") fuel assemblies in a range of 1.6-4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use.

  2. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    OpenAIRE

    Ben De Pauw; Alfredo Lamberti; Julien Ertveldt; Ali Rezayat; Katrien van Tichelen; Steve Vanlanduit; Francis Berghmans

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fr...

  3. Transactions of the second technical exchange meeting on fuel- and clad-motion diagnostics for LMFBR safety test facilities

    Energy Technology Data Exchange (ETDEWEB)

    DeVolpi, A. (comp.)

    1976-01-01

    Papers are presented which deal with diagnostic requirements and fuel motion monitoring capabilities of hodoscopes, coded aperture systems, x-ray radiography, and in-core detectors. Separate abstracts and indexing were prepared for each paper. (DG)

  4. A computational technique to identify the optimal stiffness matrix for a discrete nuclear fuel assembly model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Nam-Gyu, E-mail: nkpark@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Kyoung-Joo, E-mail: kyoungjoo@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Kyoung-Hong, E-mail: kyounghong@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Suh, Jung-Min, E-mail: jmsuh@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2013-02-15

    Highlights: ► An identification method of the optimal stiffness matrix for a fuel assembly structure is discussed. ► The least squares optimization method is introduced, and a closed form solution of the problem is derived. ► The method can be expanded to the system with the limited number of modes. ► Identification error due to the perturbed mode shape matrix is analyzed. ► Verification examples show that the proposed procedure leads to a reliable solution. -- Abstract: A reactor core structural model which is used to evaluate the structural integrity of the core contains nuclear fuel assembly models. Since the reactor core consists of many nuclear fuel assemblies, the use of a refined fuel assembly model leads to a considerable amount of computing time for performing nonlinear analyses such as the prediction of seismic induced vibration behaviors. The computational time could be reduced by replacing the detailed fuel assembly model with a simplified model that has fewer degrees of freedom, but the dynamic characteristics of the detailed model must be maintained in the simplified model. Such a model based on an optimal design method is proposed in this paper. That is, when a mass matrix and a mode shape matrix are given, the optimal stiffness matrix of a discrete fuel assembly model can be estimated by applying the least squares minimization method. The verification of the method is completed by comparing test results and simulation results. This paper shows that the simplified model's dynamic behaviors are quite similar to experimental results and that the suggested method is suitable for identifying reliable mathematical model for fuel assemblies.

  5. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  6. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    Science.gov (United States)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  7. Safety analyses for a SCWR in-pile fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Raque, M., E-mail: raque@iket.fzk.de [EnBW Kernkraft GmbH (Germany); Vasari, I., E-mail: ivan.vasari@tuev-sued.de [TUV Sud Energietechnik GmbH (Germany); Schulenberg, T., E-mail: schulenberg@kit.edu [Karlsruhe Inst. of Tech. (Germany)

    2011-07-01

    A Supercritical-Water Cooled Reactor (SCWR) test fuel element is intended to be inserted into a research reactor. The test section will be operated at temperatures and pressures above the thermodynamic critical point of water. It contains four fuel rods with a total heating power of 53 kW and it is connected with a 300 °C closed coolant loop, which is equipped with two active safety systems and a depressurization system to cool the fuel rods in case of an accident. The paper explains the physical models for numerical simulations of the safety system. Some accident sequences are analyzed exemplarily to illustrate the system performance. (author)

  8. Numerical investigation on the characteristics of two-phase flow in fuel assemblies with spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Chen, D.; Yang, Z.; Zhong, Y.; Xiao, Y.; Hu, L. [Chongqing Univ. (China). Key Lab. of Low-grade Energy Utilization Technologies and Systems

    2016-07-15

    In pressurized water reactors (PWRs), the spacer grids of the fuel assembly has significant impact on the thermal-hydraulic performance of the fuel assembly. Particularly, the spacer grids with the mixing vanes can dramatically enhance the secondary flow and have significant effect on the void distribution in the fuel assembly. In this paper, the CFD study has been carried out to analyze the effects of the spacer grid with the steel contacts, dimples and mixing vanes on the boiling two-phase flow characteristics, such as the two-phase flow field, the void distribution, and so on. Considered the influence of the boiling phase change on two-phase flow, a boiling model was proposed and applied in the CFD simulation by using the UDF (User Defined Function) method. Furthermore, in order to analyze the effects of the spacer grid with mixing vanes, the adiabatic (without boiling) two-phase flow has also been investigated as comparison with the boiling two-phase flow in the fuel assembly with spacer grids. The CFD simulation on two-phase flow in the fuel assembly with the proposed boiling model can predict the characteristics of two-phase flow better.

  9. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C{sub 1}X+C{sub 2}X{sub G}, where X and X{sub G} stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C{sub 1} and C{sub 2} are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  10. Method and jig for dismantling nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Megumi; Watahiki, Minoru.

    1989-08-30

    The object of the present inention is to extract a fuel element from a lower tie plate safely and at high efficiency by a remote control operation. That is, a forked top end of a lever of a dismantling jig is inserted between the tapered portion of a lower end plug and a lower tie plate. Then, a load is applied to the counter-lower end side of the lever by a motor. This exerts an elevating force to the fuel elements to easily release fixture between the lower end plug and the lower tie plate. Since the fuel can of fuel elements is not applied with a force by this mehtod, operation safety can be improved. (I.J.).

  11. A model to describe the anisotropic viscoplastic mechanical behavior of fresh and irradiated Zircaloy-4 fuel claddings under RIA loading conditions

    Science.gov (United States)

    Le Saux, M.; Besson, J.; Carassou, S.; Poussard, C.; Averty, X.

    2008-08-01

    This paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of cold-worked stress relieved (CWSR) Zircaloy-4 fuel claddings submitted to reactivity initiated accident (RIA) loading conditions. The model relies on a multiplicative viscoplastic formulation and reproduces strain hardening, strain rate sensitivity and plastic anisotropy of the material. It includes temperature, fluence and irradiation conditions dependences within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially obtained from the PROMETRA program, dedicated to the study of zirconium alloys under RIA loading conditions. Once calibrated, the model provides a reliable description of the mechanical behavior of the fresh and irradiated (fluence up to 10×1025 nm or burnup up to 64 GWd/tU) material within large temperature (from 20 °C up to 1100 °C) and strain rate ranges (from 3×10-4 s up to 5 s), representative of the RIA spectrum. Finally, the model is used for the finite element analysis of the hoop tensile tests performed within the PROMETRA program.

  12. Evaluation of the interfacial shear strength and residual stress of TiAlN coating on ZIRLO™ fuel cladding using a modified shear-lag model approach

    Science.gov (United States)

    Liu, Y.; Bhamji, I.; Withers, P. J.; Wolfe, D. E.; Motta, A. T.; Preuss, M.

    2015-11-01

    This paper investigates the residual stresses and interfacial shear strength of a TiAlN coating on Zr-Nb-Sn-Fe alloy (ZIRLO™) substrate designed to improve corrosion resistance of fuel cladding used in water-cooled nuclear reactors, both during normal and exceptional conditions, e.g. a loss of coolant event (LOCA). The distribution and maximum value of the interfacial shear strength has been estimated using a modified shear-lag model. The parameters critical to this analysis were determined experimentally. From these input parameters the interfacial shear strength between the TiAlN coating and ZIRLO™ substrate was inferred to be around 120 MPa. It is worth noting that the apparent strength of the coating is high (∼3.4 GPa). However, this is predominantly due to the large compressive residuals stress (3 GPa in compression), which must be overcome for the coating to fail in tension, which happens at a load just 150 MPa in excess of this.

  13. 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.)

  14. Silicon carbide composite for light water reactor fuel assembly applications

    Science.gov (United States)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  15. Silicon carbide composite for light water reactor fuel assembly applications

    Energy Technology Data Exchange (ETDEWEB)

    Yueh, Ken, E-mail: kyueh@epri.com [Fuel Reliability Program, EPRI, 1300 West WT Harris Blvd, Charlotte, NC 28262 (United States); Terrani, Kurt A., E-mail: terranika@ornl.gov [Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Rd. MS 6093, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The feasibility of using SiC{sub f}–SiC{sub m} composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  16. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  17. Clad Degradation - FEPs Screening Arguments

    Energy Technology Data Exchange (ETDEWEB)

    E. Siegmann

    2004-03-17

    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]).

  18. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  19. Preliminary Thermal Hydraulic Analyses of the Conceptual Core Models with Tubular Type Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chae, Hee Taek; Park, Jong Hark; Park, Cheol

    2006-11-15

    A new research reactor (AHR, Advanced HANARO Reactor) based on the HANARO has being conceptually developed for the future needs of research reactors. A tubular type fuel was considered as one of the fuel options of the AHR. A tubular type fuel assembly has several curved fuel plates arranged with a constant small gap to build up cooling channels, which is very similar to an annulus pipe with many layers. This report presents the preliminary analysis of thermal hydraulic characteristics and safety margins for three conceptual core models using tubular fuel assemblies. Four design criteria, which are the fuel temperature, ONB (Onset of Nucleate Boiling) margin, minimum DNBR (Departure from Nucleate Boiling Ratio) and OFIR (Onset of Flow Instability Ratio), were investigated along with various core flow velocities in the normal operating conditions. And the primary coolant flow rate based a conceptual core model was suggested as a design information for the process design of the primary cooling system. The computational fluid dynamics analysis was also carried out to evaluate the coolant velocity distributions between tubular channels and the pressure drop characteristics of the tubular fuel assembly.

  20. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  1. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  2. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  3. 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)

  4. Study of fuel element characteristic of SM and SMP (SM-PRIMA) fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A. [Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

    1999-07-01

    The paper discusses the techniques and results of reactor tests and post-reactor investigations of the SM reactor fuel elements and fuel elements developed in the process of designing the specialized PRIMA test reactor with the SM reactor fuel elements used as a prototype and which are referred to as the SMP fuel elements. The behavior of fuel elements under normal operating conditions and under deviation from normal operating conditions was studied to verify the calculation techniques, to check the calculation results during preparation of the SM reactor safety substantiation report and to estimate the possibility of using such fuel elements in other projects. During tests of fuel rods under deviation from normal operating conditions their advantages were shown over fuel elements, the components of which were produced using the Al-based alloys. (author)

  5. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    Science.gov (United States)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  6. 核燃料组件无损检测探测系统设计%Detection System for the Fuel Assembly Nondestructive Testing

    Institute of Scientific and Technical Information of China (English)

    崔尧; 张向阳; 何高魁

    2015-01-01

    燃料组件是反应堆的核心部分,在高温、高压及强中子辐射场等复杂环境条件下,燃料棒中芯块会出现肿胀、变形甚至包壳破裂,严重威胁反应堆的安全运行。为了更好地了解燃料组件在反应堆内的变化,研究高燃耗的燃料组件中燃料棒的中心空洞形成和燃料棒的变形情况,高能 X 射线无损检测是一种有效的技术手段。由于辐照后核燃料组件自身具有强放射性,探测系统设计中必须考虑减弱燃料组件自身辐射对探测采集的影响,因此组件探测系统中探测器阵列及准直器的优化设计十分必要。经过建模及相关模拟计算,得到了探测器单元最佳尺寸,优化了后准直器的结构设计,为提高燃料组件无损检测系统重建图像的质量提供帮助。%Fuel assemblies are the central components of a reactor.The core fuel pellets in the fuel pins will swell and deform and the fuel cladding may even break under the complex environment of high temperature,high pressure and intense neutron radiation field,which threats the safety of the reactor.To better understand the changes in the behavior of the fuel assembly in the reactor and study the central void formations and deformations of fuel pins in fuel assemblies to high burn-up,high-energy X-ray non-destructive testing is an effective technical means.Irradiated nuclear fuel assembly has a strong radioactivity,it is necessary to optimize the design of the detector system and the collimator to reduce the effect from gamma rays emitted from the irradiated fuel assem-bly during detection system designing phase.Through modeling,estimating and optimi-zation,the optimal size of the detector unit is obtained and the collimator design is opti-mized which can lay the foundation to improve the quality of the reconstructed images of the fuel assembly nondestructive system.

  7. Review of fuel assembly and pool thermal hydraulics for fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, Ferry, E-mail: roelofs@nrg.eu; Gopala, Vinay R.; Jayaraju, Santhosh; Shams, Afaque; Komen, Ed

    2013-12-15

    Highlights: • Literature review of fuel assembly and pool thermal hydraulics for fast reactors. • Experiments and state-of-the-art simulations. • For wire wrapped fuel assemblies RANS for complete fuel assembly is state-of-the-art, LES serves reference. • For pool thermal hydraulics, typically 5 to 20 million computational volumes are used in RANS simulations. • Gas entrainment analyses are extremely demanding as in addition they request multiphase modelling. -- Abstract: Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possible efficient use of uranium and the possibility to reduce the volume and lifetime of nuclear waste. Thermal-hydraulics is recognized as a key scientific subject in the development of such reactors. Two important challenges for the design of liquid metal fast reactors (LMFRs) are fuel assembly and pool thermal hydraulics. The heart of every nuclear reactor is the core, where the nuclear chain reaction takes place. Heat is produced in the nuclear fuel and transported to the coolant. LMFR core designs consist of many fuel assemblies which in turn consist of a large number of fuel rods. Wire wraps are commonly envisaged as spacer design in LMFR fuel assemblies. For the design and safety analyses of such reactors, simulations of the heat transport within the core are essential. The flow exiting the core is made up of the outlets of many different fuel assemblies. The liquid metal in these assemblies may be heated up to different temperatures. This leads to temperature fluctuations on various above core structures. As these temperature fluctuations may lead to thermal fatigue damage of the structures, an accurate characterization of the liquid metal flow field in the above core region is very important. This paper will provide an overview of state-of-the-art evaluations of fuel assembly and pool thermal hydraulics for LMFRs. It will show the tight interaction

  8. Development of Out-pile Test Technology for Fuel Assembly Performance Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; In, W. K.; Oh, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2007-03-15

    Out-pile tests with full scale fuel assembly are to verify the design and to evaluate the performance of the final products. HTL for the hydraulic tests and FAMeCT for mechanical/structural tests were constructed in this project. The maximum operating conditions of HTL are 30 bar, 320 .deg. C, and 500 m3/hr. This facility can perform the pressure drop test, fuel assembly uplift test, and flow induced vibration test. FAMeCT can perform the bending and vibration tests. The verification of the developed facilities were carried out by comparing the reference data of the fuel assembly which was obtained at the Westinghouse Co. The compared data showed a good coincidence within uncertainties. FRETONUS was developed for high temperature and high pressure fretting wear simulator and performance test. A performance test was conducted for 500 hours to check the integrity, endurance, data acquisition capability of the simulator. The technology of turbulent flow analysis and finite element analysis by computation was developed. From the establishments of out-pile test facilities for full scale fuel assembly, the domestic infrastructure for PWR fuel development has been greatly upgraded.

  9. Fabrication of nuclear fuel assemblies in Mexico; Fabricacion de ensambles de combustible nuclear en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Medrano B, A. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: amb@nuclear.inin.mx

    2007-07-01

    In the Pilot Production Plant of Nuclear Fuel facilities (PPFCN) located in the Nuclear Center of Mexico; its were processed approximately 1000 Kg of powder of uranium dioxide with 11 different enrichments from 0.71 up to 3.95% U-235, the pellets were encapsulated in Zircaloy tubes and armed around 300 rods of nuclear fuel for to manufacture four assembles of nuclear fuel and a DUMMY for the qualification of processes, personnel and equipment. The project beginning in 1990 with the one agreement among General Electric, Federal Commission of Electricity (CFE) and the National Institute of Nuclear Research (ININ), after building the PPFCN, to install equipment, to design the parameters of production and to qualify us as suppliers of nuclear fuel; it was begins in 1994 the production of four GE9B assemblies that surrendered to the CNLV in May, 1996. In 1998 its were loaded in the unit 1 of the CNLV, assemble them of nuclear fuel with serial numbers INI002, INI003, INI004 and INI005 with an average enrichment of 3.03% U-235, four complete operational cycles worked including the central control cell. During the works of the ninth recharge of the unit 1 of the CNLV, September 20, 2002 were removed these assemblies from the reactor core reaching a burnt of 35313 MWD/TMU. (Author)

  10. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  11. Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

    Science.gov (United States)

    Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

    2017-01-03

    Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm(2)) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

  12. Process for recycling components of a PEM fuel cell membrane electrode assembly

    Science.gov (United States)

    Shore, Lawrence [Edison, NJ

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  13. Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

    Science.gov (United States)

    Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

    2007-10-02

    A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

  14. Mechanical characterization tests of the KSMT06 fuel assembly and skeleton

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Heung Seok; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2011-12-15

    The KNF (KEPCO Nuclear Fuel) requested mechanical characterization tests of a fuel assembly and a skeleton of the KSMT fuel. The tests consisted of the lateral vibration and lateral/axial stiffness, lateral/axial impact and combined deflection tests carried out by using the FAMeCT (Fuel Assembly Mechanical Characterization Tester) in KAERI. The upper and lower core plate simulators were newly designed and manufactured because the fuel geometry of the KSMT06 was different from the KSNP type fuel assembly. In addition to this, the upper carriage was also revised with the LM guide system from the previous two guide rods system. Therefore, the axial and combined deflection tests were soundly executed. Each test was repeated twice to confirm the repeatability. The discrepancy from the repetition was small enough to be neglected. The mechanical characterization tests were accredited with the KOLAS (Korea Laboratory Accreditation Scheme) standard, and the certified test reports (lateral vibration, lateral/axial bending and lateral/axial impact) and the uncertified test report (combined deflection) were issued together with the current test result report

  15. Mechanical characterization tests of the X2-Gen fuel assembly and skeleton

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho; Kang, Heung Seok [KAERI, Daejeon (Korea, Republic of)

    2011-01-15

    The KNF (KEPCO Nuclear Fuel) requested mechanical characterization tests of a fuel assembly and a skeleton of the X2-Gen fuel. The tests consisted of the lateral vibration and lateral/axial stiffness, lateral/axial impact and combined deflection tests carried out by using the FAMeCT (Fuel Assembly Mechanical Characterization Tester) in KAERI. The upper and lower core plate simulators were newly designed and manufactured because the fuel geometry of the X2-Gen was different from the KSNP type fuel assembly. In addition to this, the upper carriage was also revised with the LM guide system from the previous two guide rods system. Therefore, the axial and combined deflection tests were soundly executed. Each test was repeated twice to confirm the repeatability. The discrepancy from the repetition was small enough to be neglected. The mechanical characterization tests were accredited with the KOLAS (Korea Laboratory Accreditation Scheme) standard, and the certified test reports (lateral vibration, lateral/axial bending and lateral/axial impact) and the uncertified test report (combined deflection) were issued together with the current test result report

  16. Rough draft of Technical Data Summary covering preparation of initial Mark IIIA plates by extrusion cladding

    Energy Technology Data Exchange (ETDEWEB)

    O`Leary, W.J.; Herries, R.R.

    1956-06-22

    This Technical Data Summary comprises the technical information that has been developed to date for the preparation by extrusion cladding of the outer plate of the contemplated Mark IIIA fuel plate bundle. The Mark IIIA fuel bundle is designed to produce plutonium at higher power levels than is possible with the Mark I or Mark VII assemblies. The Mark IIIA and Mark V designs are being developed concurrently for the same purpose. The information, however, on the Mark IIIA design is at present far more complete than for Mark V, and equipment for producing Mark IIIA in test quantities is already available. The purpose of this Technical Data Summary is to have available the facilities for cladding, by an apparently attractive process, a Mark IIIA plate.

  17. Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam-il [Korea Atomic Energy Research Institute, 1045 Daedeok Street, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)], E-mail: takni@kaeri.re.kr; Kim, Min-Hwan; Lee, Won Jae [Korea Atomic Energy Research Institute, 1045 Daedeok Street, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2008-10-15

    The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly.

  18. PEM fuel cell cost minimization using ``Design For Manufacture and Assembly`` techniques

    Energy Technology Data Exchange (ETDEWEB)

    Lomax, F.D. Jr.; James, B.D. [Directed Technologies, Inc., Arlington, VA (United States); Mooradian, R.P. [Ford Motor Co., Dearborn, MI (United States)

    1997-12-31

    Polymer Electrolyte Membrane (PEM) fuel cells fueled with direct hydrogen have demonstrated substantial technical potential to replace Internal Combustion Engines (ICE`s) in light duty vehicles. Such a transition to a hydrogen economy offers the potential of substantial benefits from reduced criteria and greenhouse emissions as well as reduced foreign fuel dependence. Research conducted for the Ford Motor Co. under a US Department of Energy contract suggests that hydrogen fuel, when used in a fuel cell vehicle (FCV), can achieve a cost per vehicle mile less than or equal to the gasoline cost per mile when used in an ICE vehicle. However, fuel cost parity is not sufficient to ensure overall economic success: the PEM fuel cell power system itself must be of comparable cost to the ICE. To ascertain if low cost production of PEM fuel cells is feasible, a powerful set of mechanical engineering tools collectively referred to as Design for Manufacture and Assembly (DFMA) has been applied to several representative PEM fuel cell designs. The preliminary results of this work are encouraging, as presented.

  19. Measuring the Multiplication of Spent Fuel Assemblies – It’s easier than you think!

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, Stephen Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-09

    This is a set of eight slides which advertise how easy it can be to measure the multiplication of a spent fuel assembly. A robust (fission chambers), rapid (under 15 minutes), direct (multiplication is measured, not photons from fission fragments) measurement of multiplication is possible.

  20. Characterization of spent fuel approved testing material--ATM-104

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

  1. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  2. Effect of assembly error of bipolar plate on the contact pressure distribution and stress failure of membrane electrode assembly in proton exchange membrane fuel cell

    Science.gov (United States)

    Liu, Dong'an; Peng, Linfa; Lai, Xinmin

    In practice, the assembly error of the bipolar plate (BPP) in a PEM fuel cell stack is unavoidable based on the current assembly process. However its effect on the performance of the PEM fuel cell stack is not reported yet. In this study, a methodology based on FEA model, "least squares-support vector machine (LS-SVM)" simulation and statistical analysis is developed to investigate the effect of the assembly error of the BPP on the pressure distribution and stress failure of membrane electrode assembly (MEA). At first, a parameterized FEA model of a metallic BPP/MEA assembly is established. Then, the LS-SVM simulation process is conducted based on the FEA model, and datasets for the pressure distribution and Von Mises stress of MEA are obtained, respectively for each assembly error. At last, the effect of the assembly error is obtained by applying the statistical analysis to the LS-SVM results. A regression equation between the stress failure and the assembly error is also built, and the allowed maximum assembly error is calculated based on the equation. The methodology in this study is beneficial to understand the mechanism of the assembly error and can be applied to guide the assembly process for the PEM fuel cell stack.

  3. A parametric study of assembly pressure, thermal expansion, and membrane swelling in PEM fuel cells

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2016-01-01

    Full Text Available Proton Exchange membrane (PEM fuel cells are still undergoing intense development, and the combination of new and optimized materials, improved product development, novel architectures, more efficient transport processes, and design optimization and integration are expected to lead to major gains in performance, efficiency, durability, reliability, manufacturability and cost-effectiveness. PEM fuel cell assembly pressure is known to cause large strains in the cell components. All components compression occurs during the assembly process of the cell, but also during fuel cell operation due to membrane swelling when absorbs water and cell materials expansion due to heat generating in catalyst layers. Additionally, the repetitive channel-rib pattern of the bipolar plates results in a highly inhomogeneous compressive load, so that while large strains are produced under the rib, the region under the channels remains approximately at its initial uncompressed state. This leads to significant spatial variations in GDL thickness and porosity distributions, as well as in electrical and thermal bulk conductivities and contact resistances (both at the ribe-GDL and membrane-GDL interfaces. These changes affect the rates of mass, charge, and heat transport through the GDL, thus impacting fuel cell performance and lifetime. In this paper, computational fluid dynamics (CFD model of a PEM fuel cell has been developed to simulate the pressure distribution inside the cell, which are occurring during fuel cell assembly (bolt assembling, and membrane swelling and cell materials expansion during fuel cell running due to the changes of temperature and relative humidity. The PEM fuel cell model simulated includes the following components; two bi-polar plates, two GDLs, and, an MEA (membrane plus two CLs. This model is used to study and analyses the effect of assembling and operating parameters on the mechanical behaviour of PEM. The analysis helped identifying critical

  4. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-05-01

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  5. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, T.A.

    1980-03-01

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation.

  6. Extension and assessment of the cladding ballooning model in the FRAP-T6 code

    Energy Technology Data Exchange (ETDEWEB)

    El-Adham, K

    1987-05-01

    The FRAP-T6 code was extended to calculate: (1) fuel surface azimuthal temperature distribution; (2) work done on cladding by internal pressure; and (3) azimuthal heat conduction in the cladding. The extensions were assessed by comparing calculated and measured cladding ballooning characteristics for four in-pile fuel rod tests. The assessment showed that the calculation of the fuel surface azimuthal temperature distribution improved the calculations of cladding ballooning. Both calculations and experimental results indicate that coplanar blockage due to cladding ballooning is unlikely during a large break LOCA.

  7. Verification of plutonium content in spent fuel assemblies using neutron self-interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard O [Los Alamos National Laboratory; Menlove, Apencer H [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2009-01-01

    The large amounts of plutonium in reactor spent fuel assemblies has led to increased research directed toward the measurement of the plutonium for safeguards verification. The high levels of fission product gamma-ray activity and curium neutron backgrounds have made the plutonium measurement difficult. We have developed a new technique that can directly measure both the {sup 235}U concentration and the plutonium fissile concentration using the intrinsic neutron emission fronl the curium in the fuel assembly. The passive neutron albedo reactivity (PNAR) method has been described previously where the curium neutrons are moderated in the surrounding water and reflect back into the fuel assembly to induce fissions in the fissile material in the assembly. The cadmium (Cd) ratio is used to separate the spontaneous fission source neutrons from the reflected thermal neutron fission reactions. This method can measure the sum of the {sup 235}U and the plutonium fissile mass, but not the separate components. Our new differential die-away self-interrogation method (DDSI) can be used to separate the {sup 235}U from the {sup 239}Pu. The method has been applied to both fuel rods and full assemblies. For fuel rods the epi-thermal neutron reflection method filters the reflected neutrons through thin Cd filters so that the reflected neutrons are from the epi-cadmium energy region. The neutron fission energy response in the epi-cadmium region is distinctly different for {sup 235}U and {sup 239}Pu. We are able to measure the difference between {sup 235}U and {sup 239}Pu by sampling the neutron induced fission rate as a function of time and multiplicity after the initial fission neutron is detected. We measure the neutron fission rate using list-mode data collection that stores the time correlations between all of the counts. The computer software can select from the data base the time correlations that include singles, doubles, and triples. The die-away time for the doubles

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

    Science.gov (United States)

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

    2017-05-01

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

  9. Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1987-11-01

    The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

  10. Clad fiber capacitor and method of making same

    Science.gov (United States)

    Tuncer, Enis

    2012-12-11

    A clad capacitor and method of manufacture includes assembling a preform comprising a ductile, electrically conductive fiber; a ductile, electrically insulating cladding positioned on the fiber; and a ductile, electrically conductive sleeve positioned over the cladding. One or more preforms are then bundled, heated and drawn along a longitudinal axis to decrease the diameter of the ductile components of the preform and fuse the preform into a unitized strand.

  11. CFD prediction of flow and phase distribution in fuel assemblies with spacers

    Energy Technology Data Exchange (ETDEWEB)

    Anglart, H.; Nylund, O. [ABB Atom AB, Vasteras (Switzerland); Kurul, N. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.

  12. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  13. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Science.gov (United States)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  14. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  15. Spent fuel dry storage technology development: thermal evaluation of isolated drywells containing spent fuel (1 kW PWR spent fuel assembly)

    Energy Technology Data Exchange (ETDEWEB)

    Unterzuber, R; Wright, J B

    1980-09-01

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site. Two PWR spent fuel assemblies having a decay heat level of approximately 1.1 kW were encapsulated inside the E-MAD Hot Bay and placed in instrumented near-surface drywell storage cells. Temperatures from the two isolated drywells and the adjacent soil have been recorded throughout the 19 month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (254{sup 0}F and 203{sup 0}F, respectively) during August 1979. Thereafter, all temperatures decreased and showed a cycling pattern which responded to seasonal atmospheric temperature changes. A computer model was utilized to predict the thermal response of the drywell. Computer predictions of the drywell temperatures and the temperatures of the surrounding soil are presented and show good agreement with the test data.

  16. Simulation of the nuclear fuel assembly drop test with LS-Dyna

    Energy Technology Data Exchange (ETDEWEB)

    Petkevich, P., E-mail: petya2306@gmail.com; Abramov, V.; Yuremenko, V.; Piminov, V.; Makarov, V.; Afanasiev, A.

    2014-04-01

    Transportation of the nuclear fuel containing objects is especially sensitive to accidental drops, as any event, affecting the fuel spacial arrangement, alters also neutron multiplication factor and can result in uncontrolled chain reaction. The latter is particularly important for nuclear fuel being immersed in water. Apart from that, fall can result in a mechanical damage of the fuel rods, which can cause environmental pollution by radionuclides. Final and intermediate fuel configurations during the accident depend on the impact velocity and the angle between falling object and the surface. Experiments cannot cover all the possible variants of drops, as it would result in their unacceptable prices. Therefore elaboration of the approaches to numerically simulate such kind of accidents is an essential step in the nuclear fuel transportation safety analysis and is the principal goal of the present research. Series of drop tests with fuel assemblies (FA) models of different complexity have been performed and numerically simulated with LS-Dyna software in order to proof the reliability of such kind of analysis. The paper contains description of the drop test experimental facility, some experimental results and their numerical simulation. It has been found that the finite element model of the FA and the material properties used for the simulation provide reliable predictions of the FA materials deformation and failure in case of accidental drops onto a rigid surface.

  17. High-level neutron-coincidence-counter (HLNCC) implementation: assay of the plutonium content of mixed-oxide fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Foley, J E; Bosler, G E

    1982-04-01

    The portable High-Level Neutron Coincidence Counter is used to assay the /sup 240/Pu-effective loading of a reference mixed-oxide fuel assembly by neutron coincidence counting. We have investigated the effects on the coincidence count rate of the total fuel loading (UO/sub 2/ + PuO/sub 2/), the fissile loading, the fuel rod diameter, and the fuel rod pattern. The coincidence count rate per gram of /sup 240/Pu-effective per centimeter is primarily dependent on the total fuel loading of the assembly; the higher the loading, the higher the coincidence count rate. Detailed procedures for the assay of mixed-oxide fuel assemblies are developed.

  18. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  19. Method for recovering catalytic elements from fuel cell membrane electrode assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

    2012-06-26

    A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

  20. Subchannel Analysis of Wire Wrapped SCWR Assembly

    OpenAIRE

    Jianqiang Shan; Henan Wang; Wei Liu; Linxing Song; Xuanxiang Chen; Yang Jiang

    2014-01-01

    Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assemb...

  1. Experience gained from carrying out ultrasonic cleaning of fuel assemblies and control and protection system assemblies in the Novovoronezh NPP unit 3

    Science.gov (United States)

    Gorburov, V. I.; Shvarov, V. A.; Vitkovskii, S. L.

    2014-02-01

    A growth of deposits on fuel assembly elements was revealed during operation of the Novovoronezh NPP Unit 3 starting from 1997. This growth caused progressive reduction of coolant flow rate through the reactor core and increase of pressure difference across the assemblies, which eventually led to the need to reduce the power unit output and then to shut down the power unit. In view of these circumstances, it was decided to develop an installation for ultrasonic cleaning of fuel assemblies. The following conclusions were drawn with regard of this installation after completion of all stages of its development, commissioning, and improvement: no detrimental effect of ultrasound on the integrity of fuel assemblies was revealed, whereas the cleaning effect on the fuel assemblies subjected to ultrasonic treatment and improvement of their thermal-hydraulic characteristics are obvious. With these measures implemented, it became possible to clean all fuel assemblies in the core in 2011, to achieve better thermal-hydraulic characteristics, and to avoid reduction of power output and off-scheduled outages of Unit 3.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

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

  3. Mechanical Property and Oxidation Behavior of ATF cladding developed in KAERI

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    To realize the coating cladding, coating material (Cr-based alloy) as well as coating technology (3D laser coating and arc ion plating combined with vacuum annealing) can be developed to meet the fuel cladding criteria. The coated Zr cladding can be produced after the optimization of coating technologies. The coated cladding sample showed the good oxidation/corrosion and adhesion properties without the spalling and/or severe interaction with the Zr alloy cladding from the various tests. Thus, it is known that the mechanical property and oxidation behavior of coated cladding concept developed in KAERI is reasonable for applying the ATF cladding in LWRs. At the present time various ATF concepts have been proposed and developing in many countries. The ATF concepts with potentially improved accident performance can be summarized to the coating cladding, Mo-Zr cladding, FeCrAl cladding, and SiCf/SiC cladding. Regarding the cladding performance, ATF cladding concepts will be evaluated with respect to the accident scenarios and normal operations of LWRs as well as to the fuel cladding fabrication.

  4. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2; Experiments of FCA assembly XVI-1 and their analyses

    Energy Technology Data Exchange (ETDEWEB)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Bando, Masaru

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author).

  5. Synchronized assembly of gold nanoparticles driven by a dynamic DNA-fueled molecular machine.

    Science.gov (United States)

    Song, Tingjie; Liang, Haojun

    2012-07-04

    A strategy for gold nanoparticle (AuNP) assembly driven by a dynamic DNA-fueled molecular machine is revealed here. In this machine, the aggregation of DNA-functionalized AuNPs is regulated by a series of toehold-mediated strand displacement reactions of DNA. The aggregation rate of the AuNPs can be regulated by controlling the amount of oligonucleotide catalyst. The versatility of the dynamic DNA-fueled molecular machine in the construction of two-component "OR" and "AND" logic gates has been demonstrated. This newly established strategy may find broad potential applications in terms of building up an "interface" that allows the combination of the strand displacement-based characteristic of DNA with the distinct assembly properties of inorganic nanoparticles, ultimately leading to the fabrication of a wide range of complex multicomponent devices and architectures.

  6. Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Granziera, M.R.; Kazimi, M.S.

    1980-05-01

    A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions.

  7. Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan

    2007-05-15

    This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.

  8. Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

    Science.gov (United States)

    Park, Y.; Yoo, J.; Huang, K.; Keiser, D. D.; Jue, J. F.; Rabin, B.; Moore, G.; Sohn, Y. H.

    2014-04-01

    Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45-345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr2 phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr2 and U10Mo, while the Mo2Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si)3Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo2Zr, and UZr2 phases.

  9. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  10. Automated assembling of single fuel cell units for use in a fuel cell stack

    Science.gov (United States)

    Jalba, C. K.; Muminovic, A.; Barz, C.; Nasui, V.

    2017-05-01

    The manufacturing of PEMFC stacks (POLYMER ELEKTROLYT MEMBRAN Fuel Cell) is nowadays still done by hand. Over hundreds of identical single components have to be placed accurate together for the construction of a fuel cell stack. Beside logistic problems, higher total costs and disadvantages in weight the high number of components produce a higher statistic interference because of faulty erection or material defects and summation of manufacturing tolerances. The saving of costs is about 20 - 25 %. Furthermore, the total weight of the fuel cells will be reduced because of a new sealing technology. Overall a one minute cycle time has to be aimed per cell at the manufacturing of these single components. The change of the existing sealing concept to a bonded sealing is one of the important requisites to get an automated manufacturing of single cell units. One of the important steps for an automated gluing process is the checking of the glue application by using of an image processing system. After bonding the single fuel cell the sealing and electrical function can be checked, so that only functional and high qualitative cells can get into further manufacturing processes.

  11. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru (Univ. of Tokyo (Japan)); Inoue, Akira (Tokyo Institute of Technology (Japan)); Miyazaki, Keiji (Osaka Univ. (Japan)); Abeta, Sadaaki (Mitsubishi, Tokyo (Japan)); Hori, Keiichi (Mitsubishi, Hyogo (Japan)); Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki

    1990-01-01

    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data.

  12. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Waseem,, E-mail: wazim_me@hotmail.com; Murtaza, Ghulam; Elahi, Nadeem

    2014-12-15

    Highlights: • Finite element model of CHASNUPP-1 fuel assembly produced, using Shell181 elements. • Non-linear contact and buckling analysis have been performed. • Structural integrity and stress measurement of fuel assembly is calculated. • Calculated stresses and deformations, are compared with test results. • Results of both studies are comparable, which validate finite element methodology. - Abstract: Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA) of Chashma Nuclear Power Plant-1 (CHASNUPP-1) at room temperature in air. Non-linear contact and buckling analyses have been performed using ANSYS 13.0, in-order to determine the FA's deformation behaviour as well as the location/values of the maximum stress intensity and stresses developed in axial direction under applied compression load of 7350 N or 1.6 g being the FA's handling load (Zhang et al., 1994). The finite element (FE) model comprises spacer grids, fuel rods, flexible contact between the fuel rods and grids’ supports system (springs and dimples) and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behaviour of the geometry is non-linear. The value of the perturbation force is related to the geometry of the model and/or the tolerance defined for the geometry. Therefore, a sensitivity study has been made to determine the appropriate value of an arbitrary perturbation load. It has been observed that FA deformation values obtained through FE analysis and experiment (SNERDI Tech Doc, 1994) under applied compression load are comparable and show linear behaviours. Therefore, it is confirmed that buckling of FA will not occur at the specified load. Moreover, the values of stresses obtained

  13. An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Fensin, Mike L [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory

    2009-01-01

    -term repository. The NDA of spent fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument

  14. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  15. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    Energy Technology Data Exchange (ETDEWEB)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

    2007-03-15

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

  16. Iterative ct reconstruction from few projections for the nondestructive post irradiation examination of nuclear fuel assemblies

    Science.gov (United States)

    Abir, Muhammad Imran Khan

    The core components (e.g. fuel assemblies, spacer grids, control rods) of the nuclear reactors encounter harsh environment due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of the nuclear power plants. The Post Irradiation Examination (PIE) can reveal information about the integrity of the elements during normal operations and off?normal events. Computed tomography (CT) is a tool for evaluating the structural integrity of elements non-destructively. CT requires many projections to be acquired from different view angles after which a mathematical algorithm is adopted for reconstruction. Obtaining many projections is laborious and expensive in nuclear industries. Reconstructions from a small number of projections are explored to achieve faster and cost-efficient PIE. Classical reconstruction algorithms (e.g. filtered back projection) cannot offer stable reconstructions from few projections and create severe streaking artifacts. In this thesis, conventional algorithms are reviewed, and new algorithms are developed for reconstructions of the nuclear fuel assemblies using few projections. CT reconstruction from few projections falls into two categories: the sparse-view CT and the limited-angle CT or tomosynthesis. Iterative reconstruction algorithms are developed for both cases in the field of compressed sensing (CS). The performance of the algorithms is assessed using simulated projections and validated through real projections. The thesis also describes the systematic strategy towards establishing the conditions of reconstructions and finds the optimal imaging parameters for reconstructions of the fuel assemblies from few projections.

  17. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    Energy Technology Data Exchange (ETDEWEB)

    R. Schreiner

    2004-10-21

    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. Shape optimization of wire-wrapped fuel assembly using Kriging metamodeling technique

    Energy Technology Data Exchange (ETDEWEB)

    Raza, Wasim [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Kim, Kwang-Yong [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of)], E-mail: kykim@inha.ac.kr

    2008-06-15

    In this work, shape optimization of a wire-wrapped fuel assembly in a liquid metal reactor has been carried out by combining a three-dimensional Reynolds-averaged Navier-Stokes analysis with the Kriging method, a well-known metamodeling technique for optimization. Sequential quadratic programming (SQP) is used to search the optimal point from the constructed metamodel. Two geometric design variables are selected for the optimization and design space is sampled using Latin Hypercube Sampling (LHS). The optimization problem has been defined as a maximization of the objective function, which is as a linear combination of heat transfer and friction loss related terms with a weighing factor. The objective function value is more sensitive to the ratio of the wire spacer diameter to the fuel rod diameter than to the ratio of the wire wrap pitch to the fuel rod diameter. The optimal values of the design variables are obtained by varying the weighting factor.

  19. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  20. Advanced manufacturing of intermediate temperature, direct methane oxidation membrane electrode assemblies for durable solid oxide fuel cell Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ITN proposes to create an innovative anode supported membrane electrode assembly (MEA) for solid oxide fuel cells (SOFCs) that is capable of long-term operation at...

  1. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

  2. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Science.gov (United States)

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

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  3. Gross gamma-ray measurements of light water reactor spent-fuel assemblies in underwater storage arrays

    Energy Technology Data Exchange (ETDEWEB)

    Moss, C.E.; Lee, D.M.

    1980-12-01

    Two gross gamma-ray detection systems have been developed for rapid measurement of spent-fuel assemblies in underwater storage racks. One system uses a scintillator as the detector and has a 2% crosstalk between a fuel assembly and an adjacent void. The other system uses an ion chamber as the detector. The measurements with both detectors correlate well with operator-declared burnup and cooling-time values.

  4. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  5. Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Serell, D.C.; Kaplan, S.

    1980-09-01

    Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

  6. Data summary report for the destructive examination of Rods G7, G9, J8, I9, and H6 from Turkey Point Fuel Assembly B17

    Energy Technology Data Exchange (ETDEWEB)

    Davis, R B; Pasupathi, V

    1981-04-01

    Destructive examination results of five spent fuel rods from a Turkey Point Unit 3 pressurized water reactor are reported. Examinations included fission gas analysis, cladding hydrogen content analysis, fuel burnup analysis, metallographic examination, autoradiography and shielded electron microprobe analysis. All rods were found to be of sound integrity with an average burnup of 27 GWd/MTU and a 0.3% fission gas release.

  7. Mechanical behaviour of membrane electrode assembly (MEA during cold start of PEM fuel cell from subzero environment temperature

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2015-01-01

    Full Text Available Durability is one of the most critical remaining issues impeding successful commercialization of broad PEM fuel cell transportation energy applications. Automotive fuel cells are likely to operate with neat hydrogen under load-following or load-levelled modes and be expected to withstand variations in environmental conditions, particularly in the context of temperature and atmospheric composition. In addition, they are also required to survive over the course of their expected operational lifetimes i.e., around 5,500 hrs, while undergoing as many as 30,000 startup/shutdown cycles. Cold start capability and survivability of proton exchange membrane fuel cells (PEM in a subzero environment temperature remain a challenge for automotive applications. A key component of increasing the durability of PEM fuel cells is studying the behaviour of the membrane electrode assembly (MEA at the heart of the fuel cell. The present work investigates how the mechanical behaviour of MEA are influenced during cold start of the PEM fuel cell from subzero environment temperatures. Full three-dimensional, non-isothermal computational fluid dynamics model of a PEM fuel cell has been developed to simulate the stresses inside the PEM fuel cell, which are occurring during fuel cell assembly (bolt assembling, and the stresses arise during fuel cell running due to the changes of temperature and relative humidity. The model is shown to be able to understand the many interacting, complex electrochemical, transport phenomena, and stresses distribution that have limited experimental data.

  8. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  9. FY04 Inspection Results for Wet Uruguay Fuel in L-Basin

    Energy Technology Data Exchange (ETDEWEB)

    VORMELKER, PHILIP

    2005-09-01

    The 2004 visual inspection of four Uruguay nuclear fuel assemblies stored in L-Basin was completed. This was the third inspection of this wet stored fuel since its arrival in the summer of 1998. Visual inspection photographs of the fuel from the previous and the recent inspections were compared and no evidence of significant corrosion was found on the individual fuel plate photographs. Fuel plates that showed areas of pitting in the cladding during the original receipt inspection were also identified during the 2004 inspection. However, a few pits were found on the non-fuel aluminum clamping plates that were not visible during the original and 2001 inspections.

  10. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  11. Sodium fast reactor fuels and materials : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  12. Direct borohydride fuel cell: Main issues met by the membrane-electrodes-assembly and potential solutions

    Science.gov (United States)

    Demirci, Umit B.

    The direct borohydride fuel cell (DBFC) is a fuel cell for which there is consensus about its promising commercial future as a portable power system. However, its development faces three main issues: the borohydride hydrolysis (issue 1) and crossover (issue 2), and the cost (issue 3). These issues are encountered by the membrane-electrodes-assembly. By a discussion around these three issues, the present paper reviews the experimental aspects. The discussion stresses on the opportunities of improvements and reviews the potential solutions that are proposed in the open literature. For each issue, the best solution seems to be a combination of improvements. The issue 1 may be solved thanks to a gold-based anode catalyst and an optimized fuel. The solution to the issue 2 may be a more efficient membrane combined with an optimized fuel and an inactive-towards-borohydride cathode catalyst like MnO 2. Regarding the issue 3, cheaper materials and better fuel use efficiency are the keys. The DBFC is still in a development phase with a small number of years of R&D invested and it appears that there are real improvement opportunities on the path of the DBFC marketing.

  13. Impedance Analysis of the Conditioning of PBI–Based Electrode Membrane Assemblies for High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Vang, Jakob Rabjerg; Andreasen, Søren Juhl;

    2013-01-01

    This work analyses the conditioning of single fuel cell assemblies based on different membrane electrode assembly (MEA) types, produced by different methods. The analysis was done by means of electrochemical impedance spectroscopy, and the changes in the fitted resistances of the all the tested...

  14. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Douglas Alan [Univ. of Florida, Gainesville, FL (United States)

    1991-01-01

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas & Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States` utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste.

  15. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

  16. Single-wall carbon nanotube-based proton exchange membrane assembly for hydrogen fuel cells.

    Science.gov (United States)

    Girishkumar, G; Rettker, Matthew; Underhile, Robert; Binz, David; Vinodgopal, K; McGinn, Paul; Kamat, Prashant

    2005-08-30

    A membrane electrode assembly (MEA) for hydrogen fuel cells has been fabricated using single-walled carbon nanotubes (SWCNTs) support and platinum catalyst. Films of SWCNTs and commercial platinum (Pt) black were sequentially cast on a carbon fiber electrode (CFE) using a simple electrophoretic deposition procedure. Scanning electron microscopy and Raman spectroscopy showed that the nanotubes and the platinum retained their nanostructure morphology on the carbon fiber surface. Electrochemical impedance spectroscopy (EIS) revealed that the carbon nanotube-based electrodes exhibited an order of magnitude lower charge-transfer reaction resistance (R(ct)) for the hydrogen evolution reaction (HER) than did the commercial carbon black (CB)-based electrodes. The proton exchange membrane (PEM) assembly fabricated using the CFE/SWCNT/Pt electrodes was evaluated using a fuel cell testing unit operating with H(2) and O(2) as input fuels at 25 and 60 degrees C. The maximum power density obtained using CFE/SWCNT/Pt electrodes as both the anode and the cathode was approximately 20% better than that using the CFE/CB/Pt electrodes.

  17. Non-fuel assembly components: 10 CFR 61.55 classification for waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Migliore, R.J.; Reid, B.D.; Fadeff, S.K.; Pauley, K.A.; Jenquin, U.P.

    1994-09-01

    This document reports the results of laboratory radionuclide measurements on a representative group of non-fuel assembly (NFA) components for the purposes of waste classification. This document also provides a methodology to estimate the radionuclide inventory of NFA components, including those located outside the fueled region of a nuclear reactor. These radionuclide estimates can then be used to determine the waste classification of NFA components for which there are no physical measurements. Previously, few radionuclide inventory measurements had been performed on NFA components. For this project, recommended scaling factors were selected for the ORIGEN2 computer code that result in conservative estimates of radionuclide concentrations in NFA components. These scaling factors were based upon experimental data obtained from the following NFA components: (1) a pressurized water reactor (PWR) burnable poison rod assembly, (2) a PVM rod cluster control assembly, and (3) a boiling water reactor cruciform control rod blade. As a whole, these components were found to be within Class C limits. Laboratory radionuclide measurements for these components are provided in detail.

  18. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  19. Modeling of Gap Closure in Uranium-Zirconium Alloy Metal Fuel - A Test Problem

    Energy Technology Data Exchange (ETDEWEB)

    Simunovic, Srdjan [ORNL; Ott, Larry J [ORNL; Gorti, Sarma B [ORNL; Nukala, Phani K [ORNL; Radhakrishnan, Balasubramaniam [ORNL; Turner, John A [ORNL

    2009-10-01

    Uranium based binary and ternary alloy fuel is a possible candidate for advanced fast spectrum reactors with long refueling intervals and reduced liner heat rating [1]. An important metal fuel issue that can impact the fuel performance is the fuel-cladding gap closure, and fuel axial growth. The dimensional change in the fuel during irradiation is due to a superposition of the thermal expansion of the fuel due to heating, volumetric changes due to possible phase transformations that occur during heating and the swelling due to fission gas retention. The volumetric changes due to phase transformation depend both on the thermodynamics of the alloy system and the kinetics of phase change reactions that occur at the operating temperature. The nucleation and growth of fission gas bubbles that contributes to fuel swelling is also influenced by the local fuel chemistry and the microstructure. Once the fuel expands and contacts the clad, expansion in the radial direction is constrained by the clad, and the overall deformation of the fuel clad assembly depends upon the dynamics of the contact problem. The neutronics portion of the problem is also inherently coupled with microstructural evolution in terms of constituent redistribution and phase transformation. Because of the complex nature of the problem, a series of test problems have been defined with increasing complexity with the objective of capturing the fuel-clad interaction in complex fuels subjected to a wide range of irradiation and temperature conditions. The abstract, if short, is inserted here before the introduction section. If the abstract is long, it should be inserted with the front material and page numbered as such, then this page would begin with the introduction section.

  20. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  1. Determination of spent nuclear fuel assembly multiplication with the differential die-away self-interrogation instrument

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Alexis C. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

    2014-09-01

    We present a novel method for determining the multiplication of a spent nuclear fuel assembly with a Differential Die-Away Self-Interrogation (DDSI) instrument. The signal, which is primarily created by thermal neutrons, is measured with four {sup 3}He detector banks surrounding a spent fuel assembly. The Rossi-alpha distribution (RAD) at early times reflects coincident events from single fissions as well as fission chains. Because of this fact, the early time domain contains information about both the fissile material and spontaneous fission material in the assembly being measured. A single exponential function fit to the early time domain of the RAD has a die-away time proportional to the spent fuel assembly (SFA) multiplication. This correlation was tested by simulating assay of 44 different SFAs with the DDSI instrument. The SFA multiplication was determined with a variance of 0.7%.

  2. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, Jr., Joe W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-06-19

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/137Cs 134Cs/154Eu, and 154Eu/137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

  3. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  4. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  5. Nuclear fuel assemblies' deformations measurement by optoelectronic methods in cooling ponds

    Science.gov (United States)

    Senchenko, E. S.; Zavyalov, P. S.; Finogenov, L. V.; Khakimov, D. R.

    2013-12-01

    Increasing the reliability and life-time of nuclear fuel is actual problems for nuclear power engineering. It takes to provide the high geometric stability of nuclear fuel assemblies (FA) under exploitation, since various factors cause FA mechanical deformation (bending and twisting). To obtain the objective information and make recommendations for the FA design improvement one have to fulfill the post reactor FA analysis. Therefore it takes measurements of the FA geometric parameters in cooling ponds of nuclear power plants. As applied to this problem we have developed and investigated the different optoelectronic methods, namely, structured light method, television and shadow ones. In this paper effectiveness of these methods has been investigated using the special experimental test stand and fulfilled researches are described. The experimental results of FA measurements by different methods and recommendation for their usage is given.

  6. Final Report on IFA-10, the first Swedish Instrumented Fuel Assembly Irradiated in HBWR, Norway

    Energy Technology Data Exchange (ETDEWEB)

    Gyllander, J.Aa.

    1967-12-15

    A final report is given on IFA-10, the first Swedish instrumented fuel assembly irradiated in HBWR. The post-irradiation data are presented and correlated with the irradiation statistics. No bowing of the bundle was observed, no equi-axed grain growth was discernible, the fission gas release was very small, and the relative dimensional changes in length and diameter were of the order of magnitude 9 x 10{sup -4} The hydride content of the can increased from 35 ppm to 65 ppm and, in the contact point of the spacer, to 180 ppm.

  7. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Tech