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Sample records for pwr fuel rod

  1. Axial gas flow in irradiated PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Dagbjartsson, S.J.; Murdock, B.A.; Owen, D.E.; MacDonald, P.E.

    1977-09-01

    Transient and steady state axial gas flow experiments were performed on six irradiated, commercial pressurized water reactor fuel rods at ambient temperature and 533 K. Laminar flow equations, as used in the FRAP-T2 and SSYST fuel behavior codes, were used with the gas flow results to calculate effective fuel rod radial gaps. The results of these analyses were compared with measured gap sizes obtained from metallographic examination of one fuel rod. Using measured gap sizes as input, the SSYST code was used to calculate pressure drops and mass fluxes and the results were compared with the experimental gas flow data.

  2. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  3. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail: andre@inb.gov.br; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail: adolpho@technotest.com.br

    2009-07-01

    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  4. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  5. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  6. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  7. A High Fidelity Multiphysics Framework for Modeling CRUD Deposition on PWR Fuel Rods

    Science.gov (United States)

    Walter, Daniel John

    Corrosion products on the fuel cladding surfaces within pressurized water reactor fuel assemblies have had a significant impact on reactor operation. These types of deposits are referred to as CRUD and can lead to power shifts, as a consequence of the accumulation of solid boron phases on the fuel rod surfaces. Corrosion deposits can also lead to fuel failure resulting from localized corrosion, where the increased thermal resistance of the deposit leads to higher cladding temperatures. The prediction of these occurrences requires a comprehensive model of local thermal hydraulic and chemical processes occurring in close proximity to the cladding surface, as well as their driving factors. Such factors include the rod power distribution, coolant corrosion product concentration, as well as the feedbacks between heat transfer, fluid dynamics, chemistry, and neutronics. To correctly capture the coupled physics and corresponding feedbacks, a high fidelity framework is developed that predicts three-dimensional CRUD deposition on a rod-by-rod basis. Multiphysics boundary conditions resulting from the coupling of heat transfer, fluid dynamics, coolant chemistry, CRUD deposition, neutron transport, and nuclide transmutation inform the CRUD deposition solver. Through systematic parametric sensitivity studies of the CRUD property inputs, coupled boundary conditions, and multiphysics feedback mechanisms, the most important variables of multiphysics CRUD modeling are identified. Moreover, the modeling framework is challenged with a blind comparison of plant data to predictions by a simulation of a sub-assembly within the Seabrook nuclear plant that experienced CRUD induced fuel failures. The physics within the computational framework are loosely coupled via an operator-splitting technique. A control theory approach is adopted to determine the temporal discretization at which to execute a data transfer from one physics to another. The coupled stepsize selection is viewed as a

  8. PWR-UO{sub 2} nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Sousa, R.V.; Pereira, C., E-mail: claubia@nuclear.ufmg.br; Silva, C.A.M.; Costa, A.L.; Veloso, M.A.F.; Oliveira, A.H. de

    2013-10-15

    Highlights: • A three-dimensional model of a PWR fuel were simulated. • Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. • Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. • Results show very good agreement with those published by OECD. -- Abstract: Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO{sub 2} nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants.

  9. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  10. 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.

  11. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  12. 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

  13. Study of heat transfer in a eccentric fuel rods in a non stop planned shutdown of a PWR type reactor; Estudo da transferencia de calor em uma vareta combustivel excentrica, num desligamento nao planejado de um reator do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: deisedy@gmail.com, E-mail: diogosb@outlook.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to conduct a case study in which the fuel pellets are displaced related to the center coating. Therefore, it will be addressed, first, the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, at a time later, you can use the program to know the fuel rod behavior and coolant channel.

  14. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  15. 压水堆燃料棒在轴向流作用下的随机振动响应研究%Random Response Analysis of PWR Fuel Rod Effect on Axial Flow

    Institute of Scientific and Technical Information of China (English)

    黄恒; 刘彤; 周跃民

    2015-01-01

    Based on random vibration theory ,the random response analysis method of PWR fuel rods under axial flow was established .The fluid force along the axial of rod was treated as a fluctuant random load ,and the mode shape method and power spectrum analysis method were used to derive the empirical formula of RMS response .This article provides a theoretical analysis method w hich does not rely on the flow induced vibration test of fuel assembly .The effects for the RMS response of fuel rods by the equivalent velocity ,turbulence intensity ,and correlation length factor were discussed .The method can meet the requirements of engineering analysis . The results show that the RMS response of fuel rods will increase with the equivalent velocity ,turbulence intensity and the correlation length factor .The response is more sensitive to the equivalent velocity and coefficient length factor changes ,and linearly with the turbulence intensity .In the operating condition of the pressurized water reactor (PWR) ,the RMS amplitude of fuel rods is about micrometers .%基于随机振动理论,建立了在轴向流作用下压水堆燃料棒随机响应的纯理论分析方法。将流体力考虑为沿燃料棒轴向位置的脉冲随机荷载,结合模态分析技术,从功率谱分析法推导出燃料棒振动均方根响应的表达式。提供了一套不依赖燃料组件流致振动实验的纯理论分析方法,重点分析了等效流速、湍流强度、相关长度系数等几个主要流场参数对结构均方根响应的影响。结果表明,本文计算模型的精度满足工程分析要求,燃料棒响应随等效流速、湍流强度和相关长度系数的增大而增大;其中响应对于等效流速和相关长度系数的变化较为敏感,而与湍流强度呈线性变化关系;在压水堆运行中的燃料棒均方根幅值约处在μm量级。

  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. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  18. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Science.gov (United States)

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki

    1992-06-01

    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  19. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  20. 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.

  1. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  2. Progress and prospects of nuclear fuel development in Japan, (2). Progress and future plan of research and development on PWR fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Yoshiaki; Abeta, Sadaaki; Aisu, Hideo; Teranishi, Tomoyuki

    1982-06-01

    13 years have elapsed since the first PWR plant started the operation in Japan, and at present, 11 PWR plants are in operation. During this period, much results of use and experience have been accumulated for the PWR fuel. The improvement and development of the fuel have been performed to meet the supply of the fuel sufficiently adaptable to the severe environment in Japan. In this paper, the evaluation of soundness and the improvement of reliability of PWR fuel made so far are reported, and the response of fuel side to long cycle operation and load following-up operation, which will be required in near future, is explained. The inspection of fuel has been performed at reactor sites for the purpose of sufficiently observing the irradiation behavior of fuel and detecting the points out of order. Effort has been exerted to perform various inspections thoroughly on total number of fuel and reflect the results to the improved design. Fuel leak scarcely occurred from the beginning, accordingly, improvement has been made to reduce the bending of fuel rods. The change of PWR fuel design, the evaluation of soundness and the improvement of reliability of PWR fuel, and the improvement for the future are reported.

  3. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  4. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  5. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  6. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    Energy Technology Data Exchange (ETDEWEB)

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  7. Stress Analysis of Single Spacer Grid Support considering Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

  8. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  9. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  10. PWR fuel in Japan; The changes and trend for hereafter

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki

    1992-07-01

    As for the PWR fuel in Japan, much efforts have been exerted aiming at the high reliability since the start of operation of Mihama No. 1 plant of Kansai Electric Power Co., Inc. At the beginning of 1970s, the fuel made by Westinghouse in USA was imported, and since then, the pursuit of the causes of troubles and the countermeasures and the domestic production of fuel have been carried out, and the improvement of design and the strengthening of quality control have been advanced. As the results, the occurrence of troubles decreased rapidly. As the fuel improvement for hereafter, the economical improvement by higher burnup, the saving and effective use of uranium resources as well as the increase of reliability are emphasized. The changes in the PWR fuel by Westinghouse, the course of improvement in the PWR fuel in Japan, the improvement against the troubles of the fuel, the improved design, the verification of the performance of the PWR fuel, the trend of development of the fuel such as the heightening of burnup, the saving and effective use of uranium resources, and the improved type pressurized water reactors are reported. (K.I.).

  11. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  12. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  13. PWR fuel in Japan; Progress and future trends

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki (Mitsubishi Heavy Industries Ltd., Tokyo (Japan))

    1994-06-01

    Twenty years ago, in the early years of the Japanese civil nuclear power programme, the fuel used was imported from Westinghouse in the USA. However, it was always intended that there would be a move towards fuel fabrication in Japan and by the end of 1993 around 10,000 Mitsubishi PWR fuel assemblies had been supplied to 21 PWRs in Japan. The highest burnup achieved so far is 46 GWd/t. Design changes to reduce abnormalities have been made, reliability is improving all the time and further improvements in burnup are being developed. This progress in PWR cores and fuel including MOX fuel in Japan is charted and future research and development is outlined. (UK).

  14. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Science.gov (United States)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (E<0.625 eV) are required as these affect the CR integrity through stresses/strains induced by coupled clad embrittlement / absorber swelling phenomena. The concept of the PSI scheme under development is to provide from validated core analysis models, the volumetric neutron source to a full core MCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  15. Approaches to analyze the bowing of German PWR fuel assemblies; Ansaetze zur Analyse des Biegeverhaltens deutscher DWR-Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    Boeke, H.; Bauer, R.; Bloemeling, F.; Lawall, R. [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2012-11-01

    The analysis of the bowing behavior of PWR fuel elements is required in case of increased fuel element deformations that have been observed during the last years. In the contribution the following issues are discussed: fuel element properties (stiffness, constructive features), influence factors (guiding tubes, spacer), load transfer and its impact. Under consideration of external boundary conditions an evaluation scheme was developed, using analysis data (control rod drop time), friction force measurements, fuel element characteristics (fuel element deformation, bowing) and their ranking, and simulation models (fluid-structure interactions). The evaluation scheme allows the definition of appropriate measures. The suitability of the methodology was demonstrated.

  16. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  17. Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng

    2012-01-01

    <正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.

  18. Lithium and boron analysis by LA-ICP-MS results from a bowed PWR rod with contact

    Directory of Open Access Journals (Sweden)

    Puranen Anders

    2017-01-01

    Full Text Available A previously published investigation of an irradiated fuel rod from the Ringhals 2 PWR, which was bowed to contact with an adjacent rod, identified a significant but highly localised thinning of the clad wall and increased corrosion. Rod fretting was deemed unlikely due to the adhering oxide covering the surfaces. Local overheating in itself was also deemed insufficient to account for the accelerated corrosion. Instead, an enhanced concentration of lithium due to conditions of local boiling was hypothesised to explain the accelerated corrosion. Studsvik has developed a hot cell coupled LA-ICP-MS (Laser Ablation Inductively Coupled Plasma Mass Spectrometer equipment that enables a flexible means of isotopic analysis of irradiated fuel and other highly active surfaces. In this work, the equipment was used to investigate the distribution of lithium (7Li and boron (11B in the outer oxide at the bow contact area. Depth profiling in the clad oxide at the opposite side of the rod to the point of contact, which is considered to have experienced normal operating conditions and which has a typical oxide thickness, evidenced levels of ∼10–20 ppm 7Li and a 11B content reaching hundreds of ppm in the outer parts of the oxide, largely in agreement with the expected range of Li and B clad oxide concentrations from previous studies. In the contact area, the 11B content was similar to the reference condition at the opposite side. The 7Li content in the outermost oxide closest to the contact was, however, found to be strongly elevated, reaching several hundred ppm. The considerable and highly localised increase in lithium content at the area of enhanced corrosion thus offers strong evidence for a case of lithium induced breakaway corrosion during power operation, when rod-to-rod contact and high enough surface heat flux results in a very local increase in lithium concentration.

  19. Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas

    2015-11-01

    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

  20. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  1. 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)

  2. Fuel Thermal Expansion (FTHEXP). [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G. A.

    1978-07-01

    A model is presented which deals with dimensional changes in LWR fuel pellets caused by changes in temperature. It is capable of dealing with any combination of UO/sub 2/ and PuO/sub 2/ in solid, liquid or mixed phase states, and includes expansion due to the solid-liquid phase change. The function FTHEXP models fuel thermal expansion as a function of temperature, fraction of PuO/sub 2/, and the fraction of fuel which is molten.

  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. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.

    1996-10-01

    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  5. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2015-10-15

    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  6. Timing analysis of PWR fuel pin failures

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Straka, M. (Halliburton NUS, Idaho Falls, ID (United States))

    1992-09-01

    Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report.

  7. Local Fuel Rod Crud Prediction Tool Applications

    Energy Technology Data Exchange (ETDEWEB)

    Krammen, Michael A.; Karoutas, Zeses E.; Wang, Guoqiang; Young, Michael Y

    2009-06-15

    A code system with attendant methods has been developed for modeling local fuel rod crud. This tool is used to perform the Crud Induced Localized Corrosion (CILC) risk assessment recommended by the EPRI crud and corrosion guidelines, which were developed in response to the INPO zero fuel failures by 2010 initiatives. The methodology is in production use. This paper will describe the range of problems the methodology has already been applied to and the especial pertinence to low duty fuel applications. The methodology begins with Computational Fluid Dynamics (CFD) computations over a fuel assembly grid span. The CFD results provide detailed relative variations in local heat transfer coefficient over the grid span. These very local relative variations are used to determine very local thermal hydraulic conditions over the entire axial length of every fuel rod in a reactor core over the life of the rod in reactor. The expansion using the local relative variations is currently accomplished with the HIDUTYDRV code. The very local thermal hydraulic conditions are combined with reactor coolant crud concentrations derived from EPRI BOA analysis as input to models for predicting very local fuel rod crud deposition. The reactor coolant crud concentrations are determined over each reactor cycle by reactor system wide crud mass balance calculations. The reactor coolant crud concentrations are used to calculate local crud thickness using mass transfer models which are a function of the local thermal conditions. The advanced crud deposition models also include models for calculating local crud dryout. Local crud deposition and crud dryout are strongly dependent on very local boiling or steaming, which are predicted through the translation of the CFD results. The local crud thickness and degree of local crud dryout are key factors in determining the margin or risk for local fuel rod cladding crud induced fuel failure. The development and first application of these methods was in

  8. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  9. Fabrication of preliminary fuel rods for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  10. 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.

  11. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  12. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  13. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  14. 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)

  15. 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.

  16. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  17. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  18. Evaluation of Fuel Performance Uncertainty in a PWR HFP RIA Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    Sensitivity and combined uncertainty studies based on the various kinds of uncertainty sources have been carried out in a PWR hot full power (HFP) condition. - Cladding inner diameter, fuel thermal conductivity, fuel thermal expansion and peak power have induced a significant impact to the fuel enthalpy and temperature. - Cladding hoop strain was strongly affected by the uncertainty parameters of cladding inner diameter, fuel thermal expansion, EPRI-1 CHF and peak power. - Above results are valid in the given analysis condition in this paper. Thereby, the analysis conditions, for example the peak linear heat rate before RIA or peak power and FWHM etc, are changed the results will be changed also. Approved analysis methodology for licensing application in the safety analysis of reactivity initiated accident (RIA) in Korea is based on a conservative approach. But newly introduced safety criteria, described in section 4.2 of NUREG-0800, tend to reduce the margins or depending on the reactor types rod failure is predicted due to the pellet-to-cladding mechanical interaction (PCMI) criteria. Thereby, licensee is trying to improve the margins by utilizing a less conservative approach.

  19. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  20. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  1. Spent nuclear fuel rods encapsulated in copper

    Energy Technology Data Exchange (ETDEWEB)

    Hanes, H.D.

    1984-04-01

    Using hot isostatic pressing, spent nuclear fuel rods and other radioactive wastes can be encapsulated in solid copper. The copper capsule which is formed is free of pores and cracks, and is highly resistant to attack by reducing ground waters. Such capsules should contain radioactive materials safely for hundreds of thousands of years in underground storage.

  2. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  3. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

  4. DUPIC fuel fabrication using spent PWR fuels at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

    2000-12-01

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter III, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

  5. DUPIC fuel fabrication using spent PWR fuel at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

    2001-09-01

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter 3/4y, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

  6. Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

  7. Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

    Energy Technology Data Exchange (ETDEWEB)

    Ferroukhi, H.; Vasiliev, A. [Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland); Dufresne, A. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Chawla, R. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Paul Scherrer Institut (Switzerland)

    2012-07-01

    Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

  8. LWR fuel rod behavior during reactor tests under loss-of-coolant conditions: Results of the FR2 in-pile tests

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.; Sepold, L.; Hofmann, P.; Petersen, C.; Schanz, G.; Zimmermann, H. (Kernforschungszentrum Karlsruhe G.m.b.H. (Germany, F.R.))

    1982-05-01

    Results of the FR2 in-pile tests on fuel rod behavior under loss-of-coolant accident (LOCA) conditions are presented. To investigate the possible influence of a nuclear environment on fuel rod failure mechanisms, unirradiated as well as irradiated (2500 to 35,000 MWd/tsub(U)) PWR-type test fuel rods were exposed to temperature transients simulating the second heatup phase of a LOCA. Loaded by internal overpressure, the cladding ballooned and ruptured. The burst data do not indicate major differences from results obtained out-of-pile with electrically heated fuel rod simulators, and do not show an influence of burnup. The fuel pellets in previously irradiated rods, already cracked during normal operation, crumbled completely in the regions with large cladding deformation. Post-test examinations revealed cladding mechanical behavior and oxidation to be comparable to out-of-pile results, with relatively little fission gas release during the transient.

  9. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: nolosesamuel@prodigy.net.mx

    2008-07-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  10. Prediction of CRUD deposition on PWR fuel using a state-of-the-art CFD-based multi-physics computational tool

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, Victor [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Kendrick, Brian K. [Theoretical Division (T-1, MS B221), Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Walter, Daniel [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Manera, Annalisa, E-mail: manera@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Secker, Jeffrey [Westinghouse Electric Company Nuclear Fuel Division, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-04-01

    In the present paper we report about the first attempt to demonstrate and assess the ability of state-of-the-art high-fidelity computational tools to reproduce the complex patterns of CRUD deposits found on the surface of operating Pressurized Water Reactors (PWRs) fuel rods. A fuel assembly of the Seabrook Unit 1 PWR was selected as the test problem. During Seabrook Cycle 5, CRUD induced power shift (CIPS) and CRUD induced localized corrosion (CILC) failures were observed. Measurements of the clad oxide thickness on both failed and non-failed rods are available, together with visual observations and the results from CRUD scrapes of peripheral rods. Blind simulations were performed using the Computational Fluid Dynamics (CFD) code STAR-CCM+ coupled to an advanced chemistry code, MAMBA, developed at Los Alamos National Laboratory. The blind simulations were then compared to plant data, which were released after completion of the simulations.

  11. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Energy Technology Data Exchange (ETDEWEB)

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  12. Numerical evaluation of flow through a 5X5 PWR rod bundle: effect of the vane arrangement in a spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    Spacer grids along the fuel assembly of Pressurized Water Reactors (PWR) maintain rod bundles arranged in a regular square configuration. The mixing vanes present in the spacer grids promote cross and swirl flow between and within the subchannels, enhancing the heat transfer performance in the grid vicinity, but also causing an adverse increase of the pressure drop in the rod bundle due the constriction on the coolant flow area. Therefore, the thermal hydraulic design of the grid must allow for both low pressure loss and high coolant mixing, which means it is important to optimize the design of the grid in relation to the mixing vane. More recently, Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently as an auxiliary tool in the development of spacer grids. The influence of some geometric characteristics of spacer grids on the flow through a rod bundle have been numerically evaluated and are still a subject of discussion. This work analyses the influence of the vanes arrangement in the spacer grid on the flow through a PWR 5 x 5 rod bundle segment. The Numerical simulations were performed with the commercial code CFX 11.0. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the k- turbulence model with scalable wall function was used. Five different vane arrangements were simulated at reactor level power and flow characteristics. The same grid and vane geometry were used in all simulations. The results of this study were divided in two parts. In the first part the presence of peripheral vanes on 5 x 5 rod bundle spacer grid tests were evaluated. The results showed that peripheral vanes should be avoided in experiments and simulations in order to

  13. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  14. Research on PWR Core Performance With MOX Fuel Loading%MOX燃料对压水堆堆芯性能影响研究

    Institute of Scientific and Technical Information of China (English)

    李小生; 靳忠敏

    2013-01-01

    Use of MOX fuel in nuclear reactors is an effective way to dispose of plutonium .A large PWR reactor core with full core loading UO 2 fuel was referenced , the reactor core physics parameters of PWR with whole and part core loading MOX fuel were calculated by using DRAGON and DONJON codes ,and the reactivity worth of control rods and boron acid solution were researched under loading MOX fuel . The results show that PWR core with MOX fuel can achieve the desired cycle length and power distribution ,but loading MOX fuel will significantly decrease the reactivity worth of control rod and boron acid solution ,moreover ,the proportion of loading MOX fuel is positive to the decrease degree of reactivity worth .%在核反应堆中使用MOX燃料是处置钚的有效方式。以大型全UO2燃料压水堆堆芯设计作为参考,使用DRAGON、DONJON程序,计算在大型压水堆中全堆芯及部分堆芯装载MOX燃料后反应堆部分物理性能指标,研究加入MOX燃料后对控制棒与硼酸溶液的反应性价值的影响。结果表明,压水堆堆芯装载各比例MOX燃料均可达到与全UO2燃料堆芯相当的循环长度,功率分布也能满足相应的安全限值要求,但采用MOX燃料会造成控制棒与硼溶液的反应性价值降低,且降低程度与MOX燃料装载比例成正相关。

  15. Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR

    Science.gov (United States)

    Wei, Guohai; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; He, Linfeng; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

  16. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    Energy Technology Data Exchange (ETDEWEB)

    Naegeli, Robert Earl

    2004-06-01

    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  17. Evaluation of a numeric procedure for flow simulation of a 5X5 PWR rod bundle with a mixing vane spacer

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5x5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the {kappa}-{epsilon} turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and

  18. 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.

  19. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  20. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  1. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Science.gov (United States)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  2. 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.

  3. Development of a program for the analysis on the free vibration of a fuel rod and its application

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seung; Yim, Jeong Sik [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Commercial Nuclear fuel burns more than 2 or three years in a core and it is essential that the fuels have a integrity without any failures during the burnup period. The factors that influence on the fuel integrity are classified as nuclear, mechanical, thermal and material factors and they are inter-related with complexity. Since the final integrity should be assured mechanically, the evaluation of the fuel rod mechanical integrity is important in a fuel design. The fuel rod for PWR is supported by spring of spacer grids to maintain its axial location and lateral space between fuel rods to get proper functions during the residence in a reactor. The long exposure duration makes the spring to be relax and loss the spring force that results in a fuel rod rattling which may cause fuel rod failure. The design criteria of the spring forces for various fuel vendors are similar each other but they are slightly different: require minimal spring force to prevent the spring from rattling at the end of life or the minimal gap between fuel rod and spring. However the spring force is relaxed due to the neutron irradiation and stress relaxation that suddenly decrease exponentially and the spring behave nonlinear by the initial spring deflection and the relaxation phenomenon. The objective of this study is to develop a finite element program to support the mechanical evaluation in view of the interaction between fuel rod and spacer spring. Here considering the spring behaviour as a function of burnup, the reaction forces of the springs are calculated by the finite element program, BEVIRA developed herein to aid the evaluation of the integrity of the fuel rod from fretting. A fuel rod is modelled as a beam to get natural frequencies and mode shapes supported by a rotational spring at each spacer spring. The results from the program are compared with previous data and those from ANSYS for the validation of the program and procedures. For the example calculation, the characteristics

  4. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  5. Electric Fuel Rod Simulator Fabrication at ORNL

    Science.gov (United States)

    Ott, Larry J.; McCulloch, Reg

    2004-02-01

    Commercial vendors could not supply the high-quality, highly instrumented electric fuel rod simulators (FRS) required for large thermal-hydraulic safety-oriented experiments at the Oak Ridge National Laboratory (ORNL) in the 1970s and early 1980s. Staff at ORNL designed, developed, and manufactured the simulators utilized in these safety experiments. Important FRS design requirements include (1) materials of construction, (2) test power requirements and availability, (3) experimental test objectives, (4) supporting thermal analyses, and (5) extensive quality control throughout all phases of FRS fabrication. This paper will present an overview of these requirements (design, analytics, and quality control) as practiced at ORNL to produce a durable high-quality FRS.

  6. Assessment of PWR fuel degradation by post-irradiation examinations and modeling in DEGRAD-1 code; Avaliacao da degradacao de combustivel PWR por exames pos-irradiacao e modelagem no codigo DEGRAD-1

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Lucki, Georgi; Silva, Jose Eduardo Rosa da; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: myrthes@ipen

    2005-07-01

    On the majority of the cases, the inquiries on primary failures and secondary in PWR fuel rods are based on results of analysis were made use of the non-destructive examination results (coolant activities monitoring, sipping tests, visual examination). The complementary analysis methodology proposed in this work includes a modeling approach to characterization of the physical effects of the individual chemistry mechanisms that constitute the incubation phase of degradation phenomenon after primary failure that are integrated in the reactor operational history under stationary operational regime, and normal power transients. The computational program called DEGRAD-1 was developed based on this modeling approach. The practical outcome of the program is to predict cladding regions susceptible to massive hydriding. The applications presented demonstrate the validity of proposed method and models by actual cases simulation, which (primary and secondary) defects positions were known and formation time was estimated. By using the modeling approach, a relationship between the hydrogen concentration in the gap and the inner cladding oxide thickness has been identified which, when satisfied, will induce massive hydriding. The novelty in this work is the integrated methodology, which supplements the traditional analysis methods (using data from non-destructive techniques) with mathematical models for the hydrogen evolution, oxidation and hydriding that include refined approaches and criteria for PWR fuel, and using the FRAPCON-3 fuel performance code as the basic tool. (author)

  7. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  8. On the Minimum Safety Factor in Elastic Buckling of Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Kyu; Kim, Jae Yong; Yoon, Kyung Ho; Lee, Young Ho; Lee, Kang Hee; Kang, Heung Seok; Song, Kun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Elastic buckling of a thin tube is an instantaneous collapse phenomenon due to an external pressure. This should be prohibited for a PWR (Pressurized Water Reactor) fuel rod. There is an engineering formula of it; however, safety factor used to be applied to the calculation results since there will be uncertainty in the parameters of the formulae such as dimensional tolerances, environmental conditions and so forth. It is a designer's responsibility to determine an appropriate safety factor that is acceptably economically conservative. Mechanical properties of a material are usually adopted from a material handbook. However, they are usually different from the measured values of the material actually used. A local dimension anomaly critically affects the elastic buckling. Conventional safety factors against the elastic buckling seemed to be large (more than 3.5). However, the reason for this is rarely found. Engineering experience may be incorporated. Therefore, it is highly necessary to propose a minimum safety factor on the elastic buckling while accommodating the above mentioned uncertainties. It is so especially for the dual cooled fuel rod since it has never been used before. The primary purpose of this work is to quantify the aforementioned uncertainties of the parameters in the elastic buckling formula, especially for an outer cladding of the currently studied dual cooled fuel rod. It is extended from the previous theoretical and experimental study

  9. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  10. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  11. Decision DGSNR/SD2/no.95/2005 Anomalies of rod clusters insertion in EDF PWR reactors; Decision DGSNR/SD2/no.95/2005 Anomalies d'insertion des grappes de commande des reacteurs a eau sous pression d'EDF

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-03-15

    Following reports of lengthening of the drop time of control rods on some PWR, in particular, with the deformation of the fuel assemblies, the Authority of Nuclear Safety asked, at the end of 2002, the operators to implement provisions of prevention and monitoring. In particular, this decision forced to carry out a measurement of the drop time of the rod clusters and prohibited to reload under rod clusters the assemblies during the last irradiation cycle. Since 2002, fuel assemblies with reinforced structure are gradually introduced allowing the limitation of the deformation under irradiation and a total improvement of the drop time. In 2004 on the favor of this favorable experience feedback, the ASN reduced the requirement. This favorable evolution continued. By the decision GGSNR/SD2/no.95/2005, the ASN authorizes the operator to charge fuel assemblies under rods during the last irradiation cycle and ends the obligation to carry out tests of drop time. (A.L.B.)

  12. 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)

  13. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  14. Treatment of defective fuel rods for interim storage

    Energy Technology Data Exchange (ETDEWEB)

    Muenchow, K.; Hummel, W. [AREVA NP GmbH, Erlangen (Germany)

    2013-07-01

    In this paper we look exclusively at the treatment of defective fuel rods for long-term dry interim storage at the nuclear power plant, in order to avoid off-site transports. AREVA has developed a technique that allows verifiably adequate drying of the defective fuel rods and reconstructs the barrier for retaining radioactive materials. This is done by individually encapsulating the defective fuel rods and achieving gas-tightness by seal welding. This guarantees the retention of radioactive materials during the storage period of at least 40 years in a transport and storage flask in an interim storage facility at site. (orig.)

  15. Gamma-ray spectroscopy on irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear], e-mail: laaterre@ipen.br

    2009-07-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  16. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  17. 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.

  18. 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.

  19. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  20. The reliability and innovation of Mitsubishi PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Abeta, S. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan); Takahashi, T.; Doi, S. [Mitsubishi Heavy Industries Ltd., Kobe (Japan)

    1997-12-31

    Mitsubishi has been making continuous efforts to improve fuel reliability for many years. Although no significant issues have occurred recently, Mitsubishi has been continuing to develop further design modifications including a variety of corrective designs to eliminate debris fretting or grid fretting. Since fuel reliability has been improved to almost zero defect levels, the emphasis has changed to reducing the amount of spent fuel and improving economics by extending burnup to the next target of 55 GWd/t and to using fuel resources more effectively by MOX fuel utilization. (author).

  1. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  2. 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.

  3. 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

  4. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  5. Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1978-10-01

    This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor (BRPR).

  6. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  7. Construction and utilization of linear empirical core models for PWR in-core fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k{sub {infinity}} profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results.

  8. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  9. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

  10. 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.

  11. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  12. Developing and analyzing long-term fuel management strategies for an advanced Small Modular PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aeoi.org.ir

    2017-03-15

    Highlights: • Comprehensive introduction and supplementary concepts as a review paper. • Developing an integrated long-term fuel management strategy for a SMR. • High reliable 3-D core modeling over fuel pins against the traditional LRM. • Verifying the expert rules of large PWRs for an advanced small PWR. • Investigating large numbers of safety parameters coherently. - Abstract: In this paper, long-term fuel management (FM) strategies are introduced and analyzed for a new advanced Pressurized Light Water Reactor (PWR) type of Small Modular Reactors (SMRs). The FM strategies are developed to be safe and practical for implementation as much as possible. Safety performances, economy of fuel, and Quality Assurance (QA) of periodic equilibrium conditions are chosen as the main goals. Flattening power density distribution over fuel pins is the major method to ensure safety performance; also maximum energy output or permissible discharging burn up indicates economy of fuel fabrication costs. Burn up effects from BOC to EOC have been traced, studied, and highly visualized in both of transport lattice cell calculations and diffusion core calculations. Long-term characteristics are searched to gain periodical equilibrium characteristics. They are fissile changes, neutron spectrum, refueling pattern, fuel cycle length, core excess reactivity, average, and maximum burn up of discharged fuels, radial Power Peaking Factors (PPF), total PPF, radial and axial power distributions, batch effects, and enrichment effects for fine regulations. Traditional linear reactivity model have been successfully simulated and adapted via fine core and burn up calculations. Effects of high burnable neutron poison and soluble boron are analyzed. Different numbers of batches via different refueling patterns have been studied and visualized. Expert rules for large type PWRs have been influenced and well tested throughout accurate equilibrium core calculations.

  13. Effects of sleeve blockages on axial velocity and intensity of turbulence in an unheated 7 x 7 rod bundle. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1976-01-01

    An experimental study is described which was performed to investigate the turbulent flow phenomena near postulated sleeve blockages in a model nuclear fuel rod bundle. The sleeve blockages were characteristic of fuel clad ''swelling'' or ''ballooning'' which could occur during loss-of-coolant accidents (LOCA) in pressurized water reactors. The study was conducted to provide information relative to the flow phenomena near postulated blockages to support detailed safety analyses of LOCAs. The results of the study are especially useful for verification of the hydraulic treatment of reactor core computer programs such as COBRA.

  14. Advanced PWR in-core fuel management with optimized gadolinia fuel designs

    Energy Technology Data Exchange (ETDEWEB)

    Berger, H.D.; Neufert, A. [Siemens AG / Power Generation KWU, Nuclear Fuel Cycle, Erlangen (Germany)

    1999-07-01

    Utilities operating LWRs require fuel assemblies and in-core fuel management service, which ensure safe, flexible and cost-effective production of electricity. With the reliability of the fuel having been always the most important requirement, advanced measures to minimize fuel cycle costs are receiving increasing attention in the light of the pressure on costs within the de-regulated power generation markets. The role of in-core fuel management in supporting the goal to minimize fuel cycle costs consists in the development of more demanding core loading strategies, i.e. in the first place more advanced low leakage loading patterns. A prerequisite for this type of loading pattern is the use of an optimized burnable absorber design. Gadolinia as integrated burnable absorber is a very effective means for limiting the critical boron concentration and power peaking factors. Siemens has accumulated extensive experience with Gd-fuel for almost 20 years with e.g. more than 5500 Gd-FA's delivered for PWRs and irradiated up to 65 MWd/kg{sub HM}. Current development efforts for optimizing Gd-fuel are focused on the reduction of the inherent penalties of today's Gd-Fa designs, i.e. reduced average FA enrichment and heavy metal content as well as residual reactivity binding. The most effective way to overcome these drawbacks is the reduction of the Gd{sub 2}O{sub 3} concentration to values of approximately 2 w/o, for which according to recent measurements of the heat conductivity of modern Gd-fuels the reduction of the fissile content in the Gd-rods is no longer necessary. Various feasibility studies have been performed to evaluate the consequences of low-Gd designs for both Siemens PWRs and Non-Siemens PWRs, for which more restrictive boundary conditions with respect to critical boron concentration and peaking factors have to be fulfilled. These studies as well as the first realization of an extended reactor cycle using a low Gd-Fa reload design confirm that the in

  15. PIE of the second fuel rod from the LOCA experiment (IFA-650.2)

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Jenssen, H.K.; Espeland, M.; Solum, N.O.

    2005-07-01

    The LOCA experiment on the second rod (IFA-650.2) a fresh, low-tin Zr-4, pressurised PWR rod was carried out in May 2004. The main objective was to produce ballooning, to determine the time to burst and to assess the material oxidation and hydriding kinetics. The rod pressure at hot conditions and peak PCT were 70 bar and 1050 C, respectively. To document the effect of the LOCA testing on the cladding, rod 2 was subjected in PIE to visual inspection, profilometry and metallography. On visual inspection the clad shows a typical balloon. In the region of max ballooning the clad shows a 35 mm long, < 20 mm burst opening. In the balloon region the outer clad diameter increased by <35% and locally the wall thickness reduction is >55%. The entire rod is covered with a black oxide layer. Below and above the split opening the continuous oxide layer was 40 to 50mum both on water and fuel side of the clad. At the locations of the upper and lower cladding thermocouples the oxide thickness was in the range 24-27 mum. Widmanstaetten structure is seen in the bulk of the clad and confirms the high temperature experienced. A some 40mum wide, hard and brittle zone with oxygen rich globular alpha-grains is found both at the outer and the inner edge of the clad in the balloon region. The zone is widest near the axial split (crack). In the clad few, arbitrary oriented hydride platelets are observed in the balloon area. (Author)

  16. New results from the NSRR experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide [Japan Atomic Research Institute, Toaki, Ibaraki (Japan)] [and others

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  17. 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 [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  18. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  19. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has

  20. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  1. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  3. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    Institute of Scientific and Technical Information of China (English)

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  4. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  5. Fuel relocation as deduced from the gas flow resistance and thermal behavior of Halden Assembly IFA-430. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Dagbjartsson, S. J.; Appelhans, T. D.; Quapp, W. J.

    1979-01-01

    The relationship of axial gas flow and fuel temperature measurements to fuel cracking and relocation occurring during the first month of irradiation of light water reactor fuel rods is discussed. Two types of fuel rod axial gas flow tests were used to determine the effective hydraulic diameter and its change during the ramping operations. Fuel centerline and off-center measurements are compared with the results of the gas flow analysis and pretest FRAP calculations.

  6. 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.

  7. Estimation and control in HTGR fuel rod fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Downing, D J; Bailey, M J

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented.

  8. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  9. Experimental fuel rod stored energy determination. STEED I project

    Energy Technology Data Exchange (ETDEWEB)

    Engman, U.; Malen, K. [Studsvik Nuclear AB, Nykoeping (Sweden)

    1999-06-01

    The objective of the STEED I (STored Energy/Enthalpy Determination) project was to evaluate an experimental method for producing accurate and reliable data concerning the stored energy in fuel rods during operation. The STEED data should provide useful information for LOCA evaluation, fuel design and thermo-mechanical modelling. Stored energy refers to the amount of heat, which at a certain time is stored within the fuel. Physical properties of the fuel that affect the quantity of stored energy are radial power profile, burnup, fuel geometry, fuel density and thermal conductivity and heat capacity of the fuel pellet, and the gas gap conductance. The quantity of stored energy is conveniently studied under transient conditions when all, or part of the stored heat is released. This work describes determination of the stored energy by evaluating scram tests. The R2 test reactor is well suited for this type of experiments, where the thermal response of different types of fuel rods can be evaluated and compared. Scrams have been performed with the intent to evaluate the fuel rod stored energy before the scram. Methods have been developed for evaluation of the stored energy from the scram response It was found that the time dependence for a large part of the heat release from the rod could be described by a single time constant. Evaluations of the time constant have been made from the data in different ways. The stored energy has been evaluated integrating the exponential decay. The integral of the exponential decay is the initial power multiplied by the time constant. This means that differences in the stored energy due to, for instance, rod properties or rod power dependence are best studied using the same time constant. The scram response was modelled with the TOODEE2 transient code. The calculations gave a time constant of about 4 s and very little power dependence. The experimental result is a time constant around 4 s. The small differences in the measurement results

  10. Angra-1 reactor core simulation with reduced diameter fuel rods; Simulacao do nucleo de Angra-1 com combustiveis de menor diametro de vareta

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano M; Faria, Eduardo F.; Sakai, Massao; Gomes, Sydney da S. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil)

    2000-07-01

    From the neutronic point of view, it is advantageous to use fuel elements with narrower rod diameter at Angra-1 PWR, since the reactivity level increases, and that happens mainly for higher enrichments than the ones used up to now. This fact is due to the higher moderator/fuel ratio, leading to a stronger neutron thermalization. In order to quantify this effect, the nodal core MEDIUM/SAV90 has been employed to simulate Angra-1 cycles from the present until the equilibrium cycle. The actual fuel element design has been maintained in this report, with exception of fuel rods diameter, reduced to 9 mm. Results have shown a higher reactivity and final burnup for the reduced diameter fuel rods, producing less waste for final disposal. However, the combined effect of higher elements reactivity and burnup made difficult the cycle-by-cycle fuel reload optimization. Preliminary results show possible advantages of using reduced diameter fuel rods in reload schemes type 'stop and go', but not being recommendable for extended cycles. (author)

  11. Test requirement for PIE of HANARO irradiated fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Lim, I. C.; Cho, Y. G

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U{sub 3}Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE.

  12. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  13. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  14. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  15. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    , inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

  16. Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

    Science.gov (United States)

    Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Karoutas, Zeses; Berndt, Markus

    2016-10-01

    Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuel rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid-structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.

  17. 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.

  18. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  19. Vibration mechanism of fuel rod in axial flow

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

    1998-08-01

    This is a review on the previous researches for the vibration of fuel rod induced by axial flow. The analysis methods are classified into three categories accordingly as the researchers postulate the vibration to be self-excited, forced and parametric; the self-excited mechanism by Burgreen and Quinn, the forced one by Reavis, Gorman, kanazawa, and S. Chen, and the parametric one by Y. Chen. Quinn supposed that the centrifugal force by flow exaggerated the natural bow in the cylinder, and the flexural force by it diminished the bow by turns; this interactive motion leaded cylinder to vibration. The supporters to the forced mechanism considered the forces arising from pressure perturbation within the boundary layers as vibrating sources. Y. Chen insisted that the cylinder could only be excited to vibration in resonance by the small oscillation of mean flow velocity. The previous studies were based on the simple boundary conditions such as hinged-hinged or fixed-fixed single span. Therefore, for the moreaccurate prediction of the fuel rod vibration in reactor, the further studies need to reflect the actual boundary conditions of the fuel rod like axial force and continuous supports by grids. (author). 25 refs.

  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 MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  1. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  2. 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).

  3. A particle assembly/constrained expansion (PACE) model for the formation and structure of porous metal oxide deposits on nuclear fuel rods in pressurized light water reactors

    Science.gov (United States)

    Brenner, Donald W.; Lu, Shijing; O'Brien, Christopher J.; Bucholz, Eric W.; Rak, Zsolt

    2015-02-01

    A new model is proposed for the structure and properties of porous metal oxide scales (aka Chalk River Unidentified Deposits (CRUD)) observed on the nuclear fuel rod cladding in Pressurized Water Reactors (PWR). The model is based on the thermodynamically-driven expansion of agglomerated octahedral nickel ferrite particles in response to pH and temperature changes in the CRUD. The model predicts that porous nickel ferrite with internal {1 1 1} surfaces is a thermodynamically stable structure under PWR conditions even when the free energy of formation of bulk nickel ferrite is positive. This explains the pervasive presence of nickel ferrite in CRUD, observed CRUD microstructures, why CRUD maintains its porosity, and variations in porosity within the CRUD observed experimentally. This model is a stark departure from decades of conventional wisdom and detailed theoretical analysis of CRUD chemistry, and defines new research directions for model validation, and for understanding and ultimately controlling CRUD formation.

  4. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    OpenAIRE

    Alroumi Fawaz; Kim Donghoon; Schow Ryan; Jevremovic Tatjana

    2016-01-01

    Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is ...

  5. Experiment on the improvement of sinterability for dry recycling nuclear fuel pellets by using simulated spent PWR fuel of high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, G. I.; Lee, Jae W.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Lee, J. W.; Yang, M. S.; Shin, W. C

    2004-09-01

    To study the fabrication characteristics of dry recycling nuclear fuel using spent PWR fuel with high burnup of 60,000 MWd/tU, the fission products of spent PWR fuel was analyzed by ORIGEN-2 code. Simulated spent PWR fuel pellets were fabricated by using UO{sub 2} powder added by the simulated fission products. The simulated dry-recycling-fuel pellets were fabricated by dry recycling fuel fabrication flow including 3 cycle treated OREOX(Oxidation and REduction of OXide fuel) process. A small amount of dopant such as TiO{sub 2}, Nb{sub 2}O{sub 5}, Li{sub 2}O are added to increase sinterability of the OREOX treated powder. As the results of experiments, the densities of sintered pellets without dopant ranged from 10.04 to 10.34 g/cm{sup 3}(94.3 to 97.1% of T.D.), the grain size of the pellets ranged from 3 to 4 {mu}m. The sintered density of the pellets with TiO{sub 2} or Nb{sub 2}O{sub 5} ranged from 10.46 to 10.32 g/cm{sup 3}(98.2 to 96.9 % of T.D.) The grain size of the pellets with TiO{sub 2}, Nb{sub 2}O{sub 5} or Li{sub 2}O ranged from 7.3 to 12.2 {mu}m.

  6. Derivation of correction factor to be applied for calculated results of PWR fuel isotopic composition by ORIGEN2 code

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan); Mochizuki, Hiroki [The Japan Research Institute Ltd., Tokyo (Japan)

    2001-11-01

    For providing conservative PWR spent fuel compositions from the view point of nuclear criticality safety, correction factors applicable for result of burnup calculation by ORIGEN2 were evaluated. Its conservativeness was verified by criticality calculations using MVP. To calculate these correction factors, analyses of spent fuel isotopic composition data were performed by ORIGEN2. Maximum or minimum value of the ratio of calculation result to experimental data was chosen as correction factor. These factors are given to each set of fuel assembly and ORIGEN2 library. They could be considered as the re-definition of recommended isotopic composition given in Nuclear Criticality Safety Handbook. (author)

  7. Multidimensional simulations of hydrides during fuel rod lifecycle

    Science.gov (United States)

    Stafford, D. S.

    2015-11-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

  8. 压水堆驱动线落棒历程计算%Calculation of Drop Course of Control Rod Assembly in PWR

    Institute of Scientific and Technical Information of China (English)

    周肖佳; 毛飞; 闵鹏; 林绍萱

    2013-01-01

    控制棒落棒性能验证是核电厂安全分析的重要部分,研制驱动线落棒历程计算程序有利于验证和改进控制棒驱动线设计。基于驱动线结构特点,分析运动组件的受力情况并进行分解,选择理论或数值方法逐一求取各分力的瞬态值,从而建立驱动线落棒历程的循环步进计算程序。利用秦山核电二期工程驱动线落棒性能试验数据对理论模型和程序计算结果进行对比验证。结果证明:所建立的驱动线落棒历程计算程序适用于压水堆驱动线系统,能正确地对运动组件落棒受力与运动历程进行模拟。%The validation of control rod drop performance is an important part of safety analysis of nuclear power plant .Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line .Based on structural features of the drive line ,the driving force on moving assembly was analyzed and decomposed ,the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method , and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved .The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR .It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA .

  9. FDD-1 System On-line Monitoring Fuel Rod Failure of Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; JISong-tao; GAOYong-guang; YINZhen-guo; HANChuan-bin

    2003-01-01

    The FDD-1 system developed by CIAE for on-line monitoring fuel rod failure of nuclear power plant consists of γ-ray detector, γ-ray spectrum analyzer, computer, and an analysis code for evaluating the status of fuel rod failure. It would be determined that the fuel rod failure occurs when a large amount of γ activity increases in the primary system measured by γ-ray detector near the CVCS.

  10. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  11. 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.

  12. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  13. Vernotte-Cattaneo approximation for heat conduction in fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Espinosa M, E. G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: gepe@xanum.uam.mx

    2009-10-15

    In this paper we explore the applicability of a fuel rod mathematical model based on the Vernotte-Cattaneo transient heat conduction as constitutive law (Non-Fourier approach) for light water reactors transient analysis. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the main steam isolated valves transient in a boiling water reactor was analyzed for different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Vernotte-Cattaneo approach can be important, while for short-times and from the engineering point of view the changes are very small. (Author)

  14. Fuel utilization improvements in a once-through PWR fuel cycle. Final report on Task 6

    Energy Technology Data Exchange (ETDEWEB)

    Dabby, D.

    1979-06-01

    In studying the position of the United States Department of Energy, Non-proliferation Alternative Systems Assessment Program, this report determines the uranium saving associated with various improvement concepts applicable to a once-through fuel cycle of a standard four-loop Westinghouse Pressurized Water Reactor. Increased discharged fuel burnup from 33,000 to 45,000 MWD/MTM could achieve a 12% U/sub 3/O/sub 8/ saving by 1990. Improved fuel management schemes combined with coastdown to 60% power, could result in U/sub 3/O/sub 8/ savings of 6%.

  15. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  16. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  17. 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.

  18. Preliminary Study on Method of Quantitative Measurement of Nuclear Fuel Rod by Neutron CT at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HE; Lin-feng; WANG; Yu; WU; Mei-mei; LIU; Yun-tao; CHEN; Dong-feng

    2015-01-01

    Neutron CT technique was applied to the quantitative measurement of the key parameters of nuclear fuel rods at China Advanced Research Reactor(CARR).The sample of dummy nuclear fuel rod was rotated in 180°range,and 900neutron projections were obtained.The 3-D neutron

  19. A MATLAB-Linked Solver to Find Fuel Depletion in a PWR, a Suggested VVER-1000 Type

    Directory of Open Access Journals (Sweden)

    F. Faghihi

    2009-01-01

    Full Text Available Coupled first-order IVPs are frequently used in many parts of engineering and sciences. We present a “solver” including three computer programs which were joint with the MATLAB software to solve and plot solutions of the first-order coupled stiff or nonstiff IVPs. Some applications related to IVPs are given here using our MATLAB-linked solver. Muon catalyzed fusion in a D-T mixture is considered as a first dynamical example of the coupled IVPs. Then, we have focused on the fuel depletion in a suggested PWR including poisons burnups (xenon-135 and samarium-149, plutonium isotopes production, and uranium depletion.

  20. 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.

  1. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  2. Studies of the UO 2-zircaloy chemical interaction and fuel rod relocation modes in a severe fuel damage accident

    Science.gov (United States)

    Shiozawa, S.; Ichikawa, M.; Fujishiro, T.

    1988-06-01

    Experiments have been conducted in the Nuclear Safety Research Reactor (NSRR) at JAERI since 1975 in order to study fuel rod failure behavior under reactivity-initiated accident conditions. Recently the experiments have been focussed on fuel behavior under simulated severe fuel damage (SFD) accident conditions. UO 2-Zircaloy reaction kinetics during very rapid transients at elevated temperatures was studied from a metallurgical point of view. Equilibrium was found to be established even in very rapid transients. The reaction rate equations developed in isothermal studies can be applied to interpret the experimental results. A fuel rod relocation criterion in connection with peak temperatures, environment conditions and initial fuel rod conditions was developed. According to the test results, fuel rod melt down due to liquefaction seems unlikely below the melting temperature of β-Zircaloy.

  3. Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deokule, A.P., E-mail: abhijit.deokule1986@gmail.com [Homi Bhabha National Institute, Trombay 400 085, Mumbai (India); Vishnoi, A.K.; Dasgupta, A.; Umasankari, K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Research Centre, Trombay 400 085, Mumbai (India)

    2014-09-15

    Highlights: • Development of 19 pin annular fuel rod cluster. • Reactor physics study of designed annular fuel rod cluster. • Thermal hydraulic study of annular fuel rod cluster. - Abstract: An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

  4. Effects of fuel relocation on reflood in a partially-blocked rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

  5. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  6. Assessment of precision gamma scanning for inspecting LWR fuel rods. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, J.R.; Barnes, B.K.; Barnes, M.L.; Hamlin, D.K.; Medina-Ortega, E.G.

    1981-07-01

    Reconstruction of the radial two-dimensional distributions of fission products using projections obtained by nondestructive gamma scanning was evaluated. The filtered backprojection algorithm provided the best reconstruction for simulated gamma-ray sources, as well as for actual irradiated fuel material. Both a low-burnup (11.5 GWd/tU) light-water reactor fuel rod and a high-burnup (179.1 GWd/tU) fast breeder reactor fuel rod were examined using this technique.

  7. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    GAO; Wei; OUYANG; Ying-gen; LI; Wei-min

    2012-01-01

    <正>Due to the deeper burnup and higher fuel swelling, fast reactor metal fuel rod using 316 stainless steel cladding, replacing the traditional zirconia cladding. The diameter of fuel rod of fast reactor is much longer than that of PWR, and the cladding of stainless steel has better ductility than zirconia cladding. Using the existing shear still will cause several aspects of problem: 1) Longer diameter of rod leads to

  8. Uncertainty analysis of spent nuclear fuel isotopics and rod internal pressure

    Science.gov (United States)

    Bratton, Ryan N.

    The bias and uncertainty in fuel isotopic calculations for a well-defined radio- chemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in the SCALE code system. Isotopic predictions are compared to measurements of fuel rod MKP109 of assembly D047 from the Calvert Cliffs Unit 1 core at three axial locations, representing a range of discharged fuel burnups. A methodology is developed which quantifies the significance of input parameter uncertainties and modeling decisions on isotopic prediction by compar- ing to isotopic measurement uncertainties. The SCALE Sampler model of the D047 assembly incorporates input parameter uncertainties for key input data such as multigroup cross sections, decay constants, fission product yields, the cladding thickness, and the power history for fuel rod MKP109. The effects of each set of input parameter uncertainty on the uncertainty of isotopic predictions have been quantified. In this work, isotopic prediction biases are identified and an investiga- tion into their sources is proposed; namely, biases have been identified for certain plutonium, europium, and gadolinium isotopes for all three axial locations. More- over, isotopic prediction uncertainty resulting from only nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle as- sembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each considered WBN1 fuel rod. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burn- able absorber (IFBA) layers is

  9. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  10. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myung Sun; Lee, Jong Po; Ju, Young Sang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-11-01

    Acoustic resonance scattering from a nuclear fuel rod in water is analyzed. A new model for the background which is attributed to the interference of reflected wave and diffracted wave is found and here named {sup t}he inherent background{sup .} The resonance spectrum of a fuel rod is obtained by subtracting the inherent background from the scattered pressure. And also analyzed are the effect of material damping of cladding tube and pellet on the resonance spectrum of a fuel rod. The propagation characteristics of circumferential waves which cause the resonances of cladding tube is produced and the appropriate resonance modes for the application to the inspection of assembled fuel rods are selected. The resonance modes are experimentally measured for pre- and post-irradiated fuel rods and the validation of the fuel rod inspection using ultrasonic resonance phenomenon is examined. And thin ultrasonic sensors accessible into the narrow interval (about 2-3mm) between assembled fuel rods are designed and manufactured. 14 refs. (Author).

  11. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    Science.gov (United States)

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  12. Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

    Energy Technology Data Exchange (ETDEWEB)

    Rowsell, David Leon [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-06-01

    This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

  13. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  14. Computer simulation of the behaviour and performance of a CANDU fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A.C. [Comison Nacional de Energia Atomica (Argentina)

    1997-07-01

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  15. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    Directory of Open Access Journals (Sweden)

    Alroumi Fawaz

    2016-01-01

    Full Text Available Control rod reactivity (worths for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

  16. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  17. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  18. 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.

  19. Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Haitao, E-mail: haitaoju@gmail.com [Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan 610041 (China); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, The University of Tokyo, Hongo, Bunkyo, Tokyo 113-8656 (Japan); Oka, Yoshiaki [Joint Department of Nuclear Energy, Waseda University, Totsukamachi, Shinjuku, Tokyo 169-8050 (Japan)

    2015-08-15

    Highlights: • The improved core of Super Fast Reactor with high power density is analyzed. • We analyzed four types of the limiting fuel rods. • The influence of Pu enrichment and compressive stress to yield strength ratio are analyzed. • The improved fuel rod design of the new core is suggested. - Abstract: A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8 W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8 W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumferential direction should

  20. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  1. Fabrication and Quality Inspection of U-10wt.%Zr Fuel Rod for Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Hwan; Song, Hoon; Oh, Seok Jin; Lee, Jung Won; Park, Jeong Yong; Lee, Chan Bock [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. Metal fuels such as U-Zr alloy have been considered as a starting driver fuel for a proto-type Gen-IV sodium cooled fast reactor (PGSFR) in Korea. To confirm the design and fabrication technologies of metallic fuels with FMS cladding for the loading of metallic fuel in PGSFR, an irradiation test will be performed in BOR-60 in Russia in 2016. In this study, U-10wt.%Zr fuel rods using low enrichment uranium (LEU) have been fabricated and inspected in quality for the fuel verification of PGSFR. Fuel slugs per melting batch without casting defects were fabricated by development of the advanced casting technology and evaluation tests. The optimal GTAW welding conditions and parameters were also established through lots of experiments. And, the qualification test carried out to prove the weld quality of end plug welding of the metallic fuel rods. The wire wrapping of metallic fuel rods for the irradiation test was successfully accomplished in KAERI. So, PGSFR fuel rods for the irradiation test in BOR-60 have been soundly fabricated in KAERI.

  2. Test Methodology of Reproducing Fuel Rod Failure by Debris Fretting Wear

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Oh Joon; Park, Nam Gyu; Kim, Jae Ik [KEPCO NF, Daejeon (Korea, Republic of)

    2015-10-15

    A test was conducted with simple debris to reproduce debris fretting wear. 68% of fuel rod cladding thickness is worn out by Inconel debris in 75 hours. The test result shows that a simple link system is useful to accommodate debris oscillation, and mid grid mixing vanes could be a source of debris forcing. Additional tests will be conducted with various debris such as wire brush, metal chip, etc which are suspected to generate actual debris fretting wear in future works. Debris fretting is one of the most common cause of the nuclear fuel rod failure. Even the most of the nuclear fuels has debris protection system, debris still cause fuel rod failure. From 1994 to 2006, debris fretting failure is around 11% of the total fuel failure. In 2006-2010, the portion of debris rises to over 13%. The total number of fuel rods failure is decreasing, but the portion of the debris fretting wear is growing with time. Therefore reproducing and identifying the mechanism of fuel rod failure by debris fretting wear is needed to improve reliability of the nuclear fuel.

  3. Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

  4. Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, J.H.

    1981-01-01

    The technical requirements for the Light Water Breeder Reactor (LWBR) fuel elements (fuel rods) imposed certain unique requirements for the low hafnium Zircaloy-4 tubing used as fuel rod cladding. This report describes, in detail, the tube manufacturing process, the product and process controls used, the inspections and tests performed, and the efforts involved in refining a commercial tube reducing process to produce tubes that would satisfy the requirements for LWBR fuel rod cladding.

  5. 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.

  6. Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation

    Directory of Open Access Journals (Sweden)

    C. Mesado

    2012-01-01

    Full Text Available In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods.

  7. Comparison of Fuel Temperature Coefficients of PWR UO{sub 2} Fuel from Monte Carlo Codes (MCNP6.1 and KENO6)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-O; Roh, Gyuhong; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a result, there was a difference within about 300-400 pcm between keff values at each enrichment due to the difference of codes and nuclear data used in the evaluations. The FTC was changed to be less negative with the increase of uranium enrichment, and it followed the form of asymptotic curve. However, it is necessary to perform additional study for investigating what factor causes the differences more than two standard deviation (2σ) among the FTCs at partial enrichment region. The interaction probability of incident neutron with nuclear fuel is depended on the relative velocity between the neutron and the target nuclei. The Fuel Temperature Coefficient (FTC) is defined as the change of Doppler effect with respect to the change in fuel temperature without any other change such as moderator temperature, moderator density, etc. In this study, the FTCs for UO{sub 2} fuel were evaluated by using MCNP6.1 and KENO6 codes based on a Monte Carlo method. In addition, the latest neutron cross-sections (ENDF/B-VI and VII) were applied to analyze the effect of these data on the evaluation of FTC, and nuclear data used in MCNP calculations were generated from the makxsf code. An evaluation of the Doppler effect and FTC for UO{sub 2} fuel widely used in PWR was conducted using MCNP6.1 and KENO6 codes. The ENDF/B-VI and VII were also applied to analyze what effect these data has on those evaluations. All cross-sections needed for MCNP calculation were produced using makxsf code. The calculation models used in the evaluations were based on the typical PWR UO{sub 2} lattice.

  8. Technical Development of the Small Fission Gas Measurement in Fuel Rods using the Laser Puncturing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heemoon; Baik, Seungje; Jin, Younggwan; Jung, Yanghong; Yoo, Boungok; Ahn, Sangbok; Yang, Yongsik; Lee, Byoungoon

    2013-12-15

    Information of fuel cladding tube and expected gas amount were obtained from fuel development department to design chamber volume and specification of laser device. Laser puncturing tests for several tubes were performed to setup power and capability. Laser puncturing tests for several tubes were performed to setup power and capability. Vacuum system with chamber was established. Additionally, QMS(Quadruple Mass Spectrometer in high vacuum state) was installed in vacuum system. The system was installed in hotcell following the preliminary test for the puncturing, pressure measuring and gas content analysis. After system test was installed in hotcell following the preliminary test for the puncturing, pressure measuring and gas content analysis. After system test was completed, SFR fuel rods were punctured to measure total gas amount and each gas content(He, Xe, Kr). The system for laser puncturing and measurement of small fission gas amout in fuel rod was designed with considering hotcell facility and fuel rod condition for first year. Chamber size, laser capability were well operated and the system showed reasonable results. In second year, QMS(Quadruple Mass Spectrometer) was installed in the system for quantitative analysis of gas contents. Thus, Laser puncturing, amount of gas measurement and gas analysis were carried out in one time. The system was activated for SFR fuel rods after installation and preliminary test. 9 SFR fuel rods were tested and produced total gas amounts and gas analysis data(He, Xe, Kr)

  9. Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hong [ORNL; Wang, Jy-An John [ORNL

    2017-01-01

    The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bending Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.

  10. 压水堆控制棒价值误差分析%Error Analysis for PWR Control Rod Integrate Worth

    Institute of Scientific and Technical Information of China (English)

    付学峰; 王磊; 郑继业; 蔡德昌; 张洪; 李冬生

    2013-01-01

    During physical startup test,the control rod integrate worth is a parameter which is most likely to be overstepped.This paper analyzes the factors which contribute to the control rod worth error and its feature.Typical example is also included.To reduce the control rod error,it is strongly suggested to create an accurate reflector model,select fuel assembly manufacture data as input and consider the control rod worth measurement method and condition when performing the calculation.To identify the main reason of control rod worth error,flux mapping test results,boron concentration and other core parameters should be analyzed comprehensively,combining with the control rod worth error distribution characteristics.%压水堆启动物理试验时,控制棒价值是比较容易超限的一个参数.本文系统分析了影响控制棒价值计算值与测量值偏差的主要因素以及各因素的影响特点、大小,并给出了部分实例分析,以期降低控制棒价值的误差,减少因控制棒价值超差对启动物理试验带来的不利影响,并在控制棒价值超差原因分析时提供帮助.分析表明,为降低控制棒价值误差,需要建立精确、合理的反射层模型,尽可能采用燃料组件的制造参数,控制棒的计算方法要考虑试验方法与工况;将注量率图试验结果、硼浓度和其他堆芯参数与控制棒价值误差分布特点相结合,进行原因查找.

  11. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    Directory of Open Access Journals (Sweden)

    Giovedi Claudia

    2016-01-01

    Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

  12. MATPRO: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Thompson, L.B. (eds.)

    1976-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Documentation and formulations that are generally semiempirical in nature are presented for uranium dioxide and mixed uranium-plutonium dioxide fuel, zircaloy cladding, gas mixture, and LWR fuel rod material properties.

  13. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  14. Standard PWR for Italy

    Energy Technology Data Exchange (ETDEWEB)

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  15. 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

  16. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements; PETER Loop. Multifunktionsversuchstand zur thermohydraulischen Untersuchung von DWR Brennelementen

    Energy Technology Data Exchange (ETDEWEB)

    Ganzmann, I.; Hille, D.; Staude, U. [AREVA NP GmbH (Germany). Materials, Fluid-Structure Interaction, Plant Life Management NTCM-G

    2009-07-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  17. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    would be important for NPP life time management purposes. In a similar way it is possible to lead estimation of EFCPO, Q - factors of coolant acoustic oscillatory circuit and PBF for any of updating NPP with PWR including NPP with supercritical parameters. Certainly, the quantitative characteristics of EFCPO, Q - factors and PBF will be various for each class of the nuclear reactor. Paper shows what operating control influences are necessary to remove EFCPO from area of resonant interaction with vibrations FR, FA etc. It is offered to use instrumentation and control systems to prevent operating of NPP at capacity level which provides increasing in amplitudes of pulsations of pressure. The increase in demand of the safety of NPP requires further increase of adequacy between a model and an object. The integrated PSB-VVER test facility is the 1:300 replica of the prototype reactor VVER with respect to power capacity and volume. The height evaluations of the test facility are the same as those of the original. The maximum power of heat released by an assembly of fuel rod simulators is 10 MW. PSB-VVER consists of four loops closed to the reactor model; the latter consists of a down comer section with the lower mixing chamber, a model of the reactor core (a channel with fuel rod simulators), a bypass of the reactor core model, and the upper mixing chamber. Each loop contains a reactor coolant pump, a steam generator, and a cold and hot pipeline. The test facility also includes a pressurizer and an ECCS consisting of three subsystems: a passive one, which incorporates four hydro accumulators and two active ones (a high-pressure ECCS and a low pressure ECCS). Test facility description, scheme and the measuring system are presented. Using such systems the transient processes have been investigated in accident with loss of coolant from the primary cooling system. The basic mathematical models for calculation of EFCPO are achieved. These models are intended for both one-phase and

  18. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  19. Research on the mechanism of formation of deposits in the fuel rod; Investigacion sobre el mecanismo de formacion de depositos en la barra combustible

    Energy Technology Data Exchange (ETDEWEB)

    Doncel, N.

    2012-11-01

    Nowadays, the interrelation between the chemistry of the coolant and the behavior of the fuel in the reactor core is considered one of the key points in the management of the reactor. Phenomena as the Axial Offset Anomaly and its association with potential Boron precipitation mechanisms in the crud deposited on the fuel have shown the necessity of an improvement in the knowledge of these mechanisms. Following this reasoning Enusa, in close collaboration with the national nuclear industry, and later with EPRI, has developed a project to investigate the chemical reactions determining the basic precipitation mechanism/dissolution of Boron in the fuel cladding. With this purpose, a test program in an specifically installation has been carried out to represent thermal conditions (sub-cooling Boiling rate) and chemicals (pH, concentration of nickel) of PWR fuel rods, with the main objective of detecting the Boron and Lithium into the crud layers. The main results of this investigation, as well as their conclusion, have contributed significantly to the general understanding of these phenomena, and will be presented in the following paper. (Author) 10 refs.

  20. Experimental Investigation on Flow-Induced Vibration of Fuel Rods in Supercritical Water Loop

    Directory of Open Access Journals (Sweden)

    Licun Wu

    2014-01-01

    Full Text Available The supercritical water-cooled reactor (SCWR is one of the most promising Generation IV reactors. In order to make the fuel qualification test for SCWR, a research plan is proposed to test a small scale fuel assembly in a supercritical water loop. To ensure the structure safety of fuel assembly in the loop, a flow-induced vibration experiment was carried out to investigate the vibration behavior of fuel rods, especially the vibration caused by leakage flow. From the experiment result, it can be found that: the vibration of rods is mainly caused by turbulence when flow rate is low. However, the effects of leakage flow become obvious as flow rate increases, which could changes the distribution of vibrational energy in spectrum, increasing the vibrational energy in high-frequency band. That is detrimental to the structure safety of fuel rods. Therefore, it is more reasonable to improve the design by using the spacers with blind hole, which can eliminate the leakage flow, to assemble the fuel rods in supercritical water loop. On the other hand, the experimental result could provide a benchmark for the theoretical studies to validate the applicability of boundary condition set for the leakage-flow-induced vibration.

  1. Radial power density distribution of MOX fuel rods in the IFA-651

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

  2. Irradiation Effects Test Series: Test IE-2. Test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C. M.; Croucher, D. W.; Ploger, S. A.; Mehner, A. S.

    1977-08-01

    The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m/sup 2/. After one hour at 68 kW/m, a power-cooling-mismatch sequence was initiated by a flow reduction at constant power. At a flow of 2550 kg/s-m/sup 2/, the onset of film boiling occurred on one rod, Rod IE-011. An additional flow reduction to 2245 kg/s-m/sup 2/ caused the onset of film boiling on the remaining three rods. Data are presented on the behavior of fuel rods during quasiequilibrium and during film boiling operation. The effects of initial gap size, cladding irradiation, rod power cycling, a rapid power increase, and sustained film boiling are discussed. These discussions are based on measured test data, preliminary postirradiation examination results, and comparisons of results with FRAP-T3 computer model calculations.

  3. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  4. 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

  5. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  6. Finite-element procedure for calculating the three-dimensional inelastic bowing of fuel rods (AWBA development program)

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S E

    1982-05-01

    An incremental finite element procedure is developed for calculating the in-pile lateral bowing of nuclear fuel rods. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the results of an axisymmetric stress analysis of the fuel rod. The effects which are taken into account in calculating the rod's lateral bowing include: (a) lateral, axial, and rotational motions and forces at the rod supports, (b) transverse gradients of temperature, fast-neutron flux, and fissioning rate, and (c) cladding circumferential wall thickness variation. The procedure developed in this report could be used to form the basis for a computer program to calculate the time-dependent bowing as a function of the fuel rod's operational and environmental history.

  7. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Ishijima, Kiyomi; Fuketa, Toyoshi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  8. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR; Evaluacion de la presencia de un absorbedor quemable en un ensamble 3x3 tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico)]. e-mail: mike_ipn_esfm@hotmail.com

    2008-07-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO{sub 2}. (Author)

  9. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  10. Fretting wear behavior of Cr-coated fuel rod for accident-tolerant fuel in flowing fluid

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    Fretting wear test of the Cr-coated fuel clading candidate have been performed in the flowing fluid condition in order to verify the reliability of Cr-coated layer on zirconium-based fuel cladding. Rod wear volume at each grid spring and dimple is dramaically increased with GTR gap even though each wear scar is not evenly distributed within a 1x1 grid cell.

  11. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  12. OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

  13. Investigation of feedback on neutron kinetics and thermal hydraulics from detailed online fuel behavior modeling during a boron dilution transient in a PWR with the two-way coupled code system DYN3D-TRANSURANUS

    Energy Technology Data Exchange (ETDEWEB)

    Holt, L., E-mail: lars.holt@tuev-sued.de [TÜV SÜD Energietechnik GmbH Baden-Württemberg, Gottlieb-Daimler-Str. 7, 70794 Filderstadt (Germany); Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany); Rohde, U.; Kliem, S.; Baier, S. [Helmholtz-Zentrum Dresden—Rossendorf, Reactor Safety Division, PO Box 510119, D-01314 Dresden (Germany); Seidl, M. [E.ON Kernkraft GmbH, Tresckowstr. 5, D-30457 Hannover (Germany); Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Macián-Juan, R. [Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany)

    2016-02-15

    Highlights: • General coupling interface was developed for the fuel performance code TRANSURANUS. • With this new tool simplified fuel behavior models in codes can be replaced. • The reactor dynamics code DYN3D was coupled to TRANSURANUS at assembly level. • The feedback from detailed online fuel behavior modeling is analyzed for reactivity initiated accident (RIA). • The thermal hydraulics can be affected strongly even in fresh fuel assemblies. - Abstract: Recently the reactor dynamics code DYN3D (including an internal fuel behavior model) was coupled to the fuel performance code TRANSURANUS at assembly level. The coupled code system applies the new general TRANSURANUS coupling interface, hence it can be used for one-way or two-way coupling. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach replaces completely the internal DYN3D fuel behavior model and transfers parameters like radial fuel temperature distribution and cladding temperature back to DYN3D. For the first time results of the coupled code system are presented for a post-critical-heat-flux heat transfer. The corresponding heat transfer regime is mostly film boiling, where the cladding temperature can rise several hundreds of degrees. The simulated boron dilution transient assumed an injection of a 36 m{sup 3} slug of under-borated coolant into a German pressurized water reactor (PWR) core initiated from a sub-critical reactor state (extreme reactivity initiated accident (RIA) conditions). The feedback from detailed fuel behavior modeling was found negligible on the neutron kinetics and thermal hydraulics during the first power rise. In a later phase of the transient, the node injected energy can differ 25 J/g, even still around 20 J/g for nodes without film boiling. Furthermore, the thermal hydraulics can be affected strongly even in fresh fuel assemblies, where film boiling

  14. Investigation of Backscatter X-ray imaging techniques for Uranium Dioxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, Timothy D [Rensselaer Polytechnic Institute (RPI); Hollenbach, Daniel F [ORNL; Shedlock, Daniel [Nucsafe, Inc.

    2011-01-01

    Radiography by Selective Detection (RSD), was investigated for its ability to determine the presence and types of defects in a UO{sub 2} fuel rod surrounded by zirconium cladding. Images created using a Monte Carlo model compared favorably with actual X-ray backscatter images from mock fuel rods. A fuel rod was modeled as a rectangular parallelepiped with zirconium cladding, and pencil beam X-ray sources of 160 kVp (79 keV avg) and 480 kVp (218 keV avg) were generated using the Monte Carlo N-Particle Transport Code to attempt to image void and palladium (Pd) defects in the interior and on the surface of the fuel pellet. It was found that the 160 kVp spectrum was unable to detect the presence of interior defects, whereas the 480 kVp spectrum detected them with both the standard and the RSD backscatter methods, though the RSD method was very inefficient. It was also found that both energy spectra were able to detect void and Pd defects on the surface using both imaging methods. Additionally, two mock fuel rods were imaged using a backscatter X-ray imaging system, one consisting of hafnium pellets in a Zircaloy-4 cladding and the other consisting of steel pellets in a Zircalloy-4 cladding which was then encased in a steel cladding (a double encapsulation configuration employed in irradiation and experiments). It was found that the system was capable of detecting individual HfO{sub 2} pellets in a Zircaloy-4 cladding and may be capable of detecting individual steel pellets in the double-encapsulated sample. It is expected that the system would also be capable of detecting individual UO{sub 2} pellets in a Zircaloy-4 cladding, though no UO{sub 2} fuel rod was available for imaging.

  15. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Young; Lee, Un Chul [Seoul National University, Seoul (Korea, Republic of)

    2011-12-15

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  16. Investigation of Minimum Film boiling Phenomena on Fuel Rods Under Blowdown Cooling Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bajorek; Michael Gawron; Timothy Etzel; Lucas Peterson

    2003-06-30

    Blowdon cooling heat transfer is an important process that occurs early in a hypothetical large break loss-of-coolant accident (LOCA) in a pressurized water reactor. During blowdown, the flow through the hot assembly is a post-critical heat flux dispersed droplet flow. The heat transfer mechanisms that occur in blowdown cooling are complex and depend on droplet and heated surface interaction. In a safety analysis, it is of considerable importance to determine the thermal-hydraulic conditions leading to the minimum film boiling temperature, Tmin. A flow boiling rig for measurement of blowdown cooling heat transfer and quench phenomena on a nuclear fuel rod simulator was designed and constructed for operation at up to 12.4 MPa. The test section consisted of a concentric annulus, with a 9.5 mm OD nuclear fuel rod simulator at the center. The rod was contained within a 0.85 mm thick, 19 mm OD 316 stainless steel tube, forming the flow channel. Two types of rods were tested; one type was sheathed with Inconel 600 while the other was clad with Zircaloy-2. Water was injected into the test section at the top of the heated length through an injection header. This header was an annular sign that fit around the fuel rod simulator and within the stainless steel tube. Small spacers aligned the injection header and prevented contract with either the heater rod or the tube. A series of small diameter holes at the bottom of the header caused the formation of droplets that became entrained with the steam flow. The test section design was such that quench would take place on the rod, and not along the channel outer annulus.

  17. CFD analysis of rewetting vertical nuclear fuel rod by dispersed fluid jet impingement

    Directory of Open Access Journals (Sweden)

    Ajoy Debbarma

    2016-09-01

    Full Text Available Numerical analysis of cooling assessment in hot vertical fuel rod is carryout using ANSYS 14.0 – CFX Solver. Rewetting is the process of re-establishment of coolants with hot surfaces. Numerical validation exercise carried out with number of turbulence and shear stress turbulence model fairly predict the experimental data and used for further investigation. In the present paper, dispersed fluid is simulating with CFX solver to investigate the flow boiling process in emergency cooling of vertical fuel rod. When coolants come in contact on the hot surface this may not initiated the wetting patch. However, this paper introduces the unique jet impingement direction to remove the heat from the hot surface. In this report, the rewetting temperature and wetting delay also described during in progress of wetting front movement in hot vertical rod.

  18. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  19. NDT of the fuel rods with artificial defect

    Energy Technology Data Exchange (ETDEWEB)

    Yang, S.Y.; Min, D.K.; Eom, S.H.; Chun, Y.B.; Min, D.K

    2000-07-01

    Non-destructive examination such as visual inspection, dimensional measurement, eddy current and gamma scanning have been carried out. The objective of this study is to evaluate the characteristics of spent fuels, and to obtain the basic technical data through the study of long term storage behavior of spent fuels. In the results of visual inspection, there is no observable effects around the part of artificial defect. And there is nothing unusual in the results of gamma scanning. Diameter and ovality the artificial defect were measured. The result obtained from this study will be used as a basic data for the study of behavior for spent fuel under the long term storage condition and the safety evaluation of spent fuel.

  20. 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

  1. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  2. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1999-08-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

  3. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Justin H. [BNF Consulting (United States); Kim, Seung Jun, E-mail: skim@lanl.gov [Mechanical and Thermal Engineering Group (AET-1), Los Alamos National Laboratory (United States); Jones, Barclay G. [Department of Nuclear Plasma Radiological Engineering, University of Illinois Urbana-Champaign (United States)

    2016-04-15

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2} can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  4. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  5. Experience of Areva in fuel services for PWR and BWR; Experiencia de Areva en servicios de combustible para PWR y BWR

    Energy Technology Data Exchange (ETDEWEB)

    Morales, I.

    2015-07-01

    AREVA being an integrated supplier of fuel assemblies has included in its strategy to develop services and solutions to customers who desire to improve the performance and safety of their fuel. These services go beyond the simple 'after sale' services that can be expected from a fuel supplier: The portfolio of AREVA includes a wide variety of services, from scientific calculations to fuel handling services in a nuclear power plant. AREVA is committed to collaborate and to propose best-in-class solutions that really make the difference for the customer, based on 40 years of Fuel design and manufacturing experience. (Author)

  6. Development of the vibration analysis technique of fuel rod and research on the methodology of fuel fretting wear analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Heung Seok; Kim, Kyung Kyu; Yoon, Hyung Hoo; Song, Ki Nam

    1998-12-01

    The FEM program has been developed to predict the natural frequencies, the FEM program has been developed to predict the natural frequencies, and mode shapes of fuel rod subjected to axial force and continuously supported by a rotational and vent spring system, and to calculate the minimum reaction forces of the spacer grid spring when the maximum vibration amplitude of fuel rod is known. This program has been verified by commercial ANSYS program and the vibration test of dummy rods in air. The test equipment were set to get the fifth modes of test rods. Partial slip problem has been studied for the analysis of fuel fretting problem. Firstly, the assumption of semi-infiniteness of the contact bodies were validated by finite element (FE) analysis. From FE results, a classical bodies were validated by finite element (FE) analysis. From FE results, aclassical theory of elasticity was utilized with regarding the problem as a plane problem. Secondly, the Mindlin-Cattaneo problem was re-evaluated, which gave the fundamental idea for developing the numerical tool for the shear traction on the contact. Shear force of sequentially-changing directions was considered and the corresponding shear traction was evaluated by extending the numerical tool for the Mindlin-Cattaneo problem.

  7. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  8. THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Croft, Stephen [Los Alamos National Laboratory; Favalli, Andrea [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory

    2012-06-19

    Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to a particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence of

  9. 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.

  10. Observation of the Isotopic Evolution of PWR Fuel Using an Antineutrino Detector

    CERN Document Server

    Bowden, N S; Dazeley, S; Svoboda, R; Misner, A; Palmer, T

    2008-01-01

    By operating an antineutrino detector of simple design during several fuel cycles, we have observed long term changes in antineutrino flux that result from the isotopic evolution of a commercial pressurized water reactor. Measurements made with simple antineutrino detectors of this kind offer an alternative means for verifying fissile inventories at reactors, as part of IAEA and other reactor safeguards regimes.

  11. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  12. Solution of a benchmark set problems for BWR and PWR reactors with UO{sub 2} and MOX fuels using CASMO-4; Solucion de un Conjunto de Problemas Benchmark para Reactores BWR y PWR con Combustible UO{sub 2} y MOX Usando CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G. [IPN, ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: mike_ipn_esfm@hotmail. com

    2007-07-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO{sub 2}) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  13. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-09-15

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on ~0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford “New Production Reactor,” later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory’s (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium

  14. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-07-25

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on {approx}0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford ''New Production Reactor'', later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory's (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of

  15. 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.

  16. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  17. On-line fuel and control rod integrity surveillance in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Sihver, L.; Larsson, I. [CHalmers Univ. of Technology, Nuclear Engineering, Gothenberg (Sweden); Loner, H. [Kernkraftwerk Leibstadt, Leibstadt (Switzerland); Grundin, A.; Helmersson, J-O.; Ledergerber, G. [Forsmarks Kraftgrupp AB, Osthammar (Sweden)

    2013-07-01

    Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of the nuclear power plant. Any actions to be taken in the event of a fuel failure during reactor operation should be based on the best available information regarding the failure and expected consequences. The detection of fuel and control rod failures in BWRs is usually performed by analyzing samples of off-gases and coolant taken with a certain time intervals, e.g. once a week or once a month. This procedure can, however, leave the failure undetected in the core for quite some time. Therefore, a sufficient improvement of the surveillance of fuel and control rods can be achieved by simultaneous measurements of He and gamma emitting noble gases on-line in the off gas system. In this paper, experiences of such measurements performed at Kernkraftwerk Leibstadt (KKL) in Switzerland and Forsmark nuclear power plant (NPP) in Sweden will be presented. (author)

  18. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  19. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  20. 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.

  1. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  2. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    Science.gov (United States)

    Foad, Basma; Takeda, Toshikazu

    2015-12-01

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO2 and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  3. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    Energy Technology Data Exchange (ETDEWEB)

    Foad, Basma [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan); Egypt Nuclear and Radiological Regulatory Authority, 3 Ahmad El Zomar St., Nasr City, Cairo, 11787 (Egypt); Takeda, Toshikazu [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan)

    2015-12-31

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  4. Feasibility study of on-line digital X-ray imaging for irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Parthoens, Y.; Gys, A. [Reactor Material Research Department, SCK-CEN, Mol (Belgium); Smolders, V. [Industrial Engineer Department, Katholieke Hogeschool Kempen, Geel (Belgium)

    2003-07-01

    At the Reactor Material Research Department of the Belgian Nuclear Research Centre SCK-CEN Xray imaging of the internal parts of irradiated fuel rods is done on silver-halide films using a 420 kV X-ray source. The replacement of the films by an on-line digital X-ray imaging system implies several advantages. Images can be evaluated instantly and source parameters can be optimized more easily. Time consuming film development is superfluous. The images can digitally be enhanced, processed, reported and archived. Within this work the feasibility of four commercial on-line digital X-ray imaging systems were studied for post-irradiation examination on fuel rods in a hot cell environment. The criteria to evaluate the systems were image quality, integration in the existing hot cell infrastructure, durability and cost price. For the evaluation and comparison of the image quality a simulation fuel rod was fabricated. Three systems suffered from lack of sensitivity, contrast and/or resolution. Only the CsI-scintillator coupled to a CCD-camera with image intensifier gave a sufficient image quality. On the other hand the image intensifiers' dimensions are difficult to integrate in the existing hot cell infrastructure. Also the durability of intensifier screens is questionable as they are susceptible to image burn. Smaller image intensifiers easier to integrate are commercial available nowadays.

  5. Mathematical modelling of friction-vibration interactions of nuclear fuel rods

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2016-06-01

    Full Text Available Nuclear fuel rods (FRs are transverselly linked to each other by three spacer grid cells at several vertical levels inside a fuel assembly (FA. Vibration of FA components, caused by the motion of FA support plates in the reactor core, generates variable contact forces between FRs and spacer grid cells. Friction effects in contact surfaces have an influence on the expected lifetime period of nuclear FA in terms of FR cladding fretting wear. This paper introduces an original approach to mathematical modelling and simulation analysis of FR nonlinear vibrations and fretting wear taking into consideration friction forces at all levels of spacer grids.

  6. Development of Application Technology of a Kagome Truss for a Fuel rod Support Structure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ki Ju; Lee, Byung Chul; Kim, Pan Su [Chonnam National University, Gwangju (Korea, Republic of)

    2010-05-15

    The purpose of this work is to design a Wire-woven Bulk Kagome (WBK) cellular metal for a fuel rod support structure of a dual cooled fuel and to fabricate test samples. Design of WBK-based support - To analyze dynamic characteristics of a support structure with WBK core under side impact. - To specify strength of WBK to be used for the support. - To design strut length and diameter of WBK. Fabrication of the test samples - To assemble WBK samples from helically formed wires. - To braze WBK samples with side straps

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

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...... uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15...

  8. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  9. A quantitative estimate on the heat transfer in cylindrical fuel rods to account for flux depression inside fuel

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mario A.B. da; Narain, Rajendra; Vasconcelos, Wagner E. de, E-mail: narain@ufpe.b, E-mail: wagner@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Dept. de Energia Nuclear

    2011-07-01

    In a nuclear reactor, the amount of power generation is limited by thermal rather than by nuclear considerations. The reactor core must be operated at a power level that the temperatures of the fuel and cladding anywhere in the core must not exceed safe limits so as to prevent from fuel element damages. Heat transfer from fuel pins can be calculated analytically by using a flat power density in the fuel pin. In actual practice, the neutron flux distribution inside fuel pins results in a smaller effective distance for the heat to be transported to the coolant. This inherent phenomenon gives rise to a heat transfer benefit in fuel pin temperatures. In this research, a quantitative estimate for transferring heat from cylindrical fuel rods is accomplished by considering a non-uniform neutron flux, which leads to a flux depression factor. This, in turn, shifts the temperature inside the fuel pin. A theoretical relationship combining the flux depression factor and a ratio of temperature gradients for uniform and non-uniform is derived, and a computational program, based on energy balance, is developed to validate the considered approximation. (author)

  10. A methodology for the evaluation of fuel rod failures under transportation accidents

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J.Y.R.; Machiels, A.J. [ANATECH, San Diego, CA (United States)]|[EPRI, Palo Alto (United States)

    2004-07-01

    Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

  11. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Lafleur, Adrienne M [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  12. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  13. Band Width of Acoustic Resonance Frequency Relatively Natural Frequency of Fuel Rod Vibration

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich; Moukhine, V.S.; Novikov, K.S.; Galivets, E.Yu. [MPEI - TU, 14, Krasnokazarmennaya str., Moscow, 111250 (Russian Federation)

    2009-06-15

    In flow induced vibrations the fluid flow is the energy source that causes vibration. Acoustic resonance in piping may lead to severe problems due to over-stressing of components or significant losses of efficiency. Steady oscillatory flow in NPP primary loop can be induced by the pulsating flow introduced by reactor circulating pump or may be set up by self-excitation. Dynamic forces generated by the turbulent flow of coolant in reactor cores cause fuel rods (FR) and fuel assembly (FA) to vibrate. Flow-induced FR and FA vibrations can generally be broken into three groups: large amplitude 'resonance type' vibrations, which can cause immediate rod failure or severe damage to the rod and its support structure, middle amplitude 'within bandwidth of resonance frequency type' vibrations responsible for more gradual wear and fatigue at the contact surface between the fuel cladding and rod support and small amplitude vibrations, 'out of bandwidth of resonance frequency type' responsible for permissible wear and fatigue at the contact surface between the fuel cladding and rod support. Ultimately, these vibration types can result in a cladding breach, and therefore must be accounted for in the thermal hydraulic design of FR and FA and reactor internals. In paper the technique of definition of quality factor (Q) of acoustic contour of the coolant is presented. The value of Q defines a range of frequencies of acoustic fluctuations of the coolant within which the resonance of oscillations of the structure and the coolant is realized. Method of evaluation of so called band width (BW) of acoustic resonance frequency is worked out and presented in the paper. BW characterises the range of the frequency of coolant pressure oscillations within which the frequency of coolant pressure oscillations matches the fuel assembly's natural frequency of vibration (its resonance frequency). Paper show the way of detuning acoustic resonance from natural

  14. A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

    Science.gov (United States)

    Fereshteh, Golian; Ali, Pazirandeh; Saeed, Mohammadi

    2015-06-01

    In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor, the Fokker-Planck equation (FPE) governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper. Besides, the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found. As for the results, the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found. Also, different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod, i.e. Krypton, Xenon, Iodine, Bromine, Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution. The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution. The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap. It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.

  15. Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

    Full Text Available An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

  16. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y. [Hitachi-GE Nuclear Energy, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-0073 (Japan); Makigami, T. [Tokyo Electric Power Company Inc., 1-1-3, Uchisaiwai-cho, Chiyoda-ku, Tokyo, 100-0011 (Japan)

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  17. CFD Validation Benchmark Dataset for Natural Convection in Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Smith, Barton; Jones, Kyle

    2016-11-01

    The present study provide CFD validation benchmark data for coupled fluid flow/convection heat transfer on the exterior of heated rods arranged in a 2 × 2 array. The rod model incorporates grids with swirling veins to resemble a nuclear fuel bundle. The four heated aluminum rods are suspended in an open-circuit wind tunnel. Boundary conditions (BCs) are measured and uncertainties calculated to provide all quantities necessary to successfully conduct a CFD validation exercise. System response quantities (SRQs) are measured for comparing the simulation output to the experiment. Stereoscopic Particle Image Velocimetry (SPIV) is used to non-intrusively measure 3-component velocity fields. A through-plane measurement is used for the inflow while laser sheet planes aligned with the flow direction at several downstream locations are used for system response quantities. Two constant heat flux rod surface conditions are presented (400 W/m2 and 700 W/m2) achieving a peak Rayleigh number of 1010 . Uncertainty for all measured variables is reported. The boundary conditions, system response, and all material properties are now available online for download. The U.S. Department of Energy Nuclear Engineering University Program provided the funding for these experiments under Grant 00128493.

  18. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  19. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  20. 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.

  1. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  2. Measurement of Fresh Fuel Rods to Demonstrate Compliance with Criticality Safety Limits

    Energy Technology Data Exchange (ETDEWEB)

    Miko, David K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-03

    In order to operate TA-66 as a radiological facility with the quantity of nuclear material required to fulfil its mission, a criticality safety evaluation was required. This evaluation defined the control parameters for operations at the facility. The resulting evaluation for TA-66 placed limits on the amount of SNM, as well as other materials such as beryllium. In addition, there is a limit on the number of uranium fuel rods allowed subject to enrichment, outer diameter, and overall length restrictions. The enrichments for the rods to be shipped to TA-66 were documented in LA-UR-13-23581, but the outer diameter and length were not documented. This report provides this information.

  3. Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

    Science.gov (United States)

    Riglet-Martial, Ch.; Sercombe, J.; Lamontagne, J.; Noirot, J.; Roure, I.; Blay, T.; Desgranges, L.

    2016-11-01

    The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that irradiated fuels (of nominal stoichiometry close to 2.000) under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U4O9 type in its cold section, of lower temperature. The parameters governing the oxidation states of UO2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

  4. Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G.A. (comp.)

    1978-02-01

    The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  5. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  6. Determination of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

    Energy Technology Data Exchange (ETDEWEB)

    Casagranda, Albert [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2016-05-01

    An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

  7. Analysis of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

    Energy Technology Data Exchange (ETDEWEB)

    Casagranda, Albert [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2016-06-01

    An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

  8. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

    2009-07-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  9. 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.)

  10. Open and closed fuel cycle of HWR and PWR. How large is the high-level radioactive wastes repository; Ciclos de combustible abierto y cerrado con HWR y PWR. ?Cuanto mas grande es el repositorio de residuos radiactivos de alta actividad?

    Energy Technology Data Exchange (ETDEWEB)

    Bevilacqua, Arturo M. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

    1996-07-01

    A conceptual analysis was carried out on the size of a high-level wastes (HLW) repository for the waste arising from once-through and closed fuel cycles with (HLW) and PWR. The mass, the activity and thermal loading was calculated with the ORIGEN2.1 computer code for the spent fuel and for the high-level liquid wastes. It was considered a minimum burnup of 7.000 MW.d/t U and 33.000 MW.d/t U for HWR and PWR respectively, cooling times of 20 and 55 years, reprocessing recovery ratios of 99% and 99.7% and a total electricity production of 81.6 GW(e).a. It was concluded that the cooling time is the most important repository size reproduction parameter for the closed cycles. On the other hand, the spent fuel mass for the once-through cycles does not depend on the cooling time what prevents repository size reduction once a cooling time of 55 years is reached. The repository size reduction in the case of HWR is larger than in the case of PWR, owing to the larger fuel mass required to produce the specific electricity amount. (author)

  11. In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.

    1980-01-01

    In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, which has been observed in all tests with previously irradiated rods, needs further investigation.

  12. The Source Term Calculation and Analysis of PWR Spent Fuel%压水堆乏燃料源项计算与分析

    Institute of Scientific and Technical Information of China (English)

    苏卓; 邹树梁; 于涛; 谢金森

    2011-01-01

    In spent fuel reprocessing plant,the highly radioactive environment will have a certain radiation damage on monitoring equipments.Therefore,the using life of the equipments will be affected and the system reliability reduced.In this article,we use ORIGEN2 program to calculate the components of PWR fuel,and obtain a group of data about the important radionuclides in spent fuel components,such like radioactivity,photon energy spectra etc.The calculated results are accurate and credible,and it can provide initial source term data for shield design of electronic monitoring equipment in the first-side processing of spent fuel.%乏燃料后处理车间的高放射性环境会对监测设备产生一定的辐照损伤,影响其使用寿命,降低系统可靠性.本文使用ORIGEN2程序对压水堆燃料组件进行计算,得出一组乏燃料组件中重要核素的放射性活度、光子能谱等数据,计算结果准确可信,可为乏燃料首端处理中电子监测设备的屏蔽设计提供初始源项数据.

  13. Synthesis and Analysis of Alpha Silicon Carbide Components for Encapsulation of Fuel Rods and Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kevin M. McHugh; John E. Garnier; George W. Griffith

    2011-09-01

    The chemical, mechanical and thermal properties of silicon carbide (SiC) along with its low neutron activation and stability in a radiation field make it an attractive material for encapsulating fuel rods and fuel pellets. The alpha phase (6H) is particularly stable. Unfortunately, it requires very high temperature processing and is not readily available in fibers or near-net shapes. This paper describes an investigation to fabricate a-SiC as thin films, fibers and near-net-shape products by direct conversion of carbon using silicon monoxide vapor at temperatures less than 1700 C. In addition, experiments to nucleate the alpha phase during pyrolysis of polysilazane, are also described. Structure and composition were characterized using scanning electron microscopy, energy dispersive spectroscopy and X-ray diffraction. Preliminary tensile property analysis of fibers was also performed.

  14. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    Science.gov (United States)

    Holcombe, Scott; Andersson, Peter; Svärd, Staffan Jacobsson; Hallstadius, Lars

    2016-11-01

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for

  15. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  16. Process Management Development for Quality Monitoring on Resistance Weldment of Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Na, Tae Hyung; Yang, Kyung Hwan; Kim, In Kyu [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    The current, welding force, and displacement are displayed on the indicator during welding. However, real-time quality control is not performed. Due to the importance of fuel rod weldment, many studies on welding procedures have been conducted. However, there are not enough studies regarding weldment quality evaluation. On the other hand, there are continuous studies on the monitoring and control of welding phenomena. In resistance welding, which is performed in a very short time, it is important to find the process parameters that well represent the weld zone formation and the welding process. In his study, Gould attempted to analyze melt zone formation using the finite difference method. Using the artificial neural network, Javed and Sanders, Messler Jr et al., Cho and Rhee, Li and Gong et al. estimated the size of the melt zone by mapping a nonlinear functional relation between the weldment and the electrode head movement, which is a typical welding process parameter. Applications of the artificial intelligence method include fuzzy control using electrode displacement, fuzzy control using the optimal power curve, neural network control using the dynamic resistance curve, fuzzy adaptive control using the optimal electrode curve, etc. Therefore, this study induced quality factors for the real-time quality control of nuclear fuel rod end plug weldment using instantaneous dynamic resistance (IDR), which incorporates the instantaneous value of secondary current and voltage of the transformer, and using instantaneous dynamic force (IDF), obtained real-time during welding.

  17. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myoung Seon; Joo, Young Sang; Jung, Hyun Kyu; Cheong, Yong Moo

    1997-02-01

    The scattering of plane acoustic waves normally incident on a multilayered cylindrical shell has been formulated using the global matrix approach. And a simple way to formulate the non-resonant background component in the field scattered by an empty elastic shell has been found. This is to replace the surface admittance for the shell with the zero-frequency limit of the surface admittance for the analogous fluid shell (i.e., the shear wave speed in the elastic shell is set to zero). It has been shown that the background thus obtained is exact and applicable to shells of arbitrary thickness and material makeup, and over all frequencies and mode numbers. This way has been also applied to obtain the expressions of the backgrounds for multilayered shells. The resonant ultrasound spectroscopy system has been constructed to measure the resonance spectrum of a single fuel rod. The leak-defective fuel rod detection system of a laboratory scale has been also constructed. Particularly, all techniques and processes necessary for manufacturing the ultrasonic probe of thin (1.2 mm) strip type have been developed. (author). 38 refs., 34 figs.

  18. Experimental and numerical study on lead-bismuth heat transfer in a fuel rod simulator

    Science.gov (United States)

    Ma, Weimin; Karbojian, Aram; Hollands, Thorsten; Koch, Marco K.

    2011-08-01

    As a task of the EU project IP EUROTRANS towards development of an Accelerator Driven System (ADS) dedicated to the transmutation of long-lived fission products, experiments and simulations were performed on the TALL test facility at KTH to investigate thermal hydraulics along a single fuel rod simulator cooled by lead-bismuth eutectic (LBE). The fuel rod simulator is concentrically inserted in a tube, so that an annular channel is formed for LBE flow. This paper presents the measured temperature profiles in the annular channel, and the comparisons with the simulation results of the CFX code. The primary objective is to help understanding the LBE heat transfer characteristics and qualifying the turbulence and heat transfer modeling for LBE application. The quantitative comparison between the calculated and measured temperatures of the LBE indicates that the simulation underestimates the experiment at most radial and axial positions. Finally the uncertainties in measurement and the deficiency in turbulence models resulting in such a disagreement were discussed, which will be directive and beneficial to future work in the field.

  19. Fuel rod model based on Non-Fourier heat conduction equation

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, Mexico DF., CP 09340 (Mexico)], E-mail: gepe@xanum.uam.mx; Espinosa-Martinez, E-G. [Retorno Quebec 6, Col. Burgos de Cuernavaca 62580, Temixco, Mor. (Mexico)

    2009-05-15

    In this paper we explore the applicability of a fuel rod mathematical model based on Non-Fourier transient heat conduction as constitutive law for the Light Water Reactors transient analysis (LWRs). In the classical theory of diffusion, Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the Main Steam Isolated Valves (MSIV) transient in a Boiling Water Reactor (BWR) was analyzed by different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Non-Fourier approach can be important, while for short-times and from the engineering point of view the changes are very small. Some results from transient calculations are examined.

  20. Thermo-Mechanical Analysis of Coated Particle Fuel Experiencing a Fast Control Rod Ejection Transient

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Brian Boer; Abderrafi M. Ougouag

    2010-10-01

    A rapid increase of the temperature and the mechanical stress is expected in TRISO coated particle fuel that experiences a fast Total Control Rod Ejection (CRE) transient event. During this event the reactor power in the pebble bed core increases significantly for a short time interval. The power is deposited instantly and locally in the fuel kernel. This could result in a rapid increase of the pressure in the buffer layer of the coated fuel particle and, consequently, in an increase of the coating stresses. These stresses determine the mechanical failure probability of the coatings, which serve as the containment of radioactive fission products in the Pebble Bed Reactor (PBR). A new calculation procedure has been implemented at the Idaho National Laboratory (INL), which analyzes the transient fuel performance behavior of TRISO fuel particles in PBRs. This early capability can easily be extended to prismatic designs, given the availability of neutronic and thermal-fluid solvers. The full-core coupled neutronic and thermal-fluid analysis has been modeled with CYNOD-THERMIX. The temperature fields for the fuel kernel and the particle coatings, as well as the gas pressures in the buffer layer, are calculated with the THETRIS module explicitly during the transient calculation. Results from this module are part of the feedback loop within the neutronic-thermal fluid iterations performed for each time step. The temperature and internal pressure values for each pebble type in each region of the core are then input to the PArticle STress Analysis (PASTA) code, which determines the particle coating stresses and the fraction of failed particles. This paper presents an investigation of a Total Control Rod Ejection (TCRE) incident in the 400 MWth Pebble Bed Modular reactor design using the above described calculation procedure. The transient corresponds to a reactivity insertion of $3 (~2000 pcm) reaching 35 times the nominal power in 0.5 seconds. For each position in the core

  1. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  2. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  3. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

    Directory of Open Access Journals (Sweden)

    Amir Zacarias Mesquita

    2013-01-01

    Full Text Available In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF. At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout. This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN. It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool.

  4. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Wilson

    2001-02-08

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  5. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth D. Wright

    1997-07-30

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  6. Elastic analysis of thermal gradient bowing in rod-type fuel elements subjected to axial thrust (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Newman, J.B.

    1968-01-01

    Thermal radient bowing of rod type fuel elements can be analyzed in terms of the deflections of a precurved beam. The fundamental aspects of an analysis of axially compressed multispan beams are given. Elasticity of supports in both axial and transverse directions is considered; the technique is applicable to problems in which the axial thrust depends on the transverse deflection as well as problems with prescribed axial thrust. The formulas presented constitute the theory for a computer program of broad applicability, not only in the analysis of fuel rod bowing, but also to almost any multispan beam, particularly when the effects of axial loads cannot be neglected. 17 references. (NSA 22: 22866)

  7. Numerical Prediction of Dual-Cooled Annular Fuel Temperature During Control Rod Ejection Accident in OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chan Eun; In, Wang Kee; Yang, Soo Hyung; Chun, Tae Hyun; Song, Kun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    A dual-cooled annular fuel concept for a light water reactor has been introduced by MIT for a significant amount of reactor power uprate. MIT proposed a 13x13 annular fuel array replacing the 17x17 solid fuel in the Westinghouse 4-loop plant, which could increase the core power up to 50% with the considerable changes in the major reactor components. The Korea Atomic Energy Research Institute (KAERI) is also conducting a research to develop a dual-cooled fuel for its employment in an optimized pressurized water reactor in Korea, OPR1000. The dual-cooled fuel for the OPR1000 is targeted to increase the reactor power by 20% as well as reduce the fuel-pellet temperature by more than 30% without a change to the reactor components other than the fuel. Numerous technical tasks exist for assessing the applicability of the dual cooled annular fuel to the power uprate in the OPR1000. One of the important tasks is to evaluate the performance of the annular fuel during the design basis events. Particularly, the fuel temperature and the peak cladding temperature (PCT) are the important variables during the control rod ejection accident (REA), since the rod averaged fuel enthalpy should be lower than its safety limit. The fuel enthalpy is known to largely depend on the fuel temperature. This paper presents the predictions of the fuel and peak cladding temperatures during the REA. A general-purpose structural code, ABAQUS-6.8 and a computational fluid dynamics code, ANSYS CFX-11.0 were used to perform the numerical analysis of a heat transfer in the annular fuel as well as the solid fuel. The numerical predictions of the fuel maximum temperature (FMT) and PCT are compared against those predicted by a best-estimate system transient analysis code, MARS.

  8. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    Science.gov (United States)

    Ratnayake, Ruwan Kumara

    While best estimate computer codes provide the licensing basis for nuclear power facilities, they also serve as analytical tools in overall plant and component design procedures. An ideal best estimate code would comprise of universally applicable mechanistic models for all its components. However, due to the limited understanding in these specific areas, many of the models and correlations used in these codes reflect high levels of empiricism. As a result, the use of such models is strictly limited to the range of parameters within which the experiments have been conducted. Disagreements between best estimate code predictions and experimental results are often explained by the mechanistic inadequacies of embedded models. Significant mismatches between calculated and experimental critical power values are common observations in the analyses of Boiling Water Reactors (BWR). Based on experimental observations and calculations, these mismatches are attributed to the additional entrainment and deposition caused by spacer grids in BWR fuel assemblies. In COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid); a state of the art industrial best estimate code, these disagreements are hypothesized to occur due the absence of an appropriate spacer grid model. In this thesis, development of a suitably detailed spacer grid model and integrating it to COBRA-TF is documented. The new spacer grid model is highly mechanistic so that the applicability of it is not seriously affected by geometric variations in different spacer grid designs. COBRA-TF (original version) simulations performed on single tube tests and BWR rod bundles with spacer grids showed that single tube predictions were more accurate than those of the rod bundles. This observation is understood to arise from the non-availability of a suitable spacer grid model in COBRA-TF. Air water entrainment experiments were conducted in a test section simulating two adjacent BWR sub channels to visualize the flow behavior at

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-30

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

  11. Two-dimensional thermal analysis of a fuel rod by finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Rhayanne Y.N.; Silva, Mario A.B. da; Lira, Carlos A.B. de O., E-mail: ryncosta@gmail.com, E-mail: mabs500@gmail.com, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamaento de Energia Nuclear

    2015-07-01

    In a nuclear reactor, the amount of power generation is limited by thermal and physic limitations rather than by nuclear parameters. The operation of a reactor core, considering the best heat removal system, must take into account the fact that the temperatures of fuel and cladding shall not exceed safety limits anywhere in the core. If such considerations are not considered, damages in the fuel element may release huge quantities of radioactive materials in the coolant or even core meltdown. Thermal analyses for fuel rods are often accomplished by considering one-dimensional heat diffusion equation. The aim of this study is to develop the first paper to verify the temperature distribution for a two-dimensional heat transfer problem in an advanced reactor. The methodology is based on the Finite Volume Method (FVM), which considers a balance for the property of interest. The validation for such methodology is made by comparing numerical and analytical solutions. For the two-dimensional analysis, the results indicate that the temperature profile agree with expected physical considerations, providing quantitative information for the development of advanced reactors. (author)

  12. Fuel performance improvement program: description and characterization of HBWR Series H-2, H-3, and H-4 test rods

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Barner, J.O.; Welty, R.K.

    1980-03-01

    The fabrication process and as-built characteristics of the HBWR Series H-2 and H-3 test rods, as well as the three packed-particle (sphere-pac) rods in HBWR Series H-4 are described. The HBWR Series H-2, H-3, and H-4 tests are part of the irradiation test program of the Fuel Performance Improvement Program. Fifteen rods were fabricated for the three test series. Rod designs include: (1) a reference dished pellet design incorporating chamfered edges, (2) a chamfered, annular pellet design combined with graphite-coated cladding, and (3) a sphere-pac design. Both the annular-coated and sphere-pac designs include internal pressurization using helium.

  13. Structure Optimization Design of the Electronically Controlled Fuel Control Rod System in a Diesel Engine

    Directory of Open Access Journals (Sweden)

    Hui Jin

    2015-01-01

    Full Text Available Poor ride comfort and shorter clutch life span are the key factors restricting the commercialization of automated manual transmission (AMT. For nonelectrically controlled engines or AMT where cooperative control between the engine and the transmission is not realizable, applying electronically controlled fuel control rod systems (ECFCRS is an effective way to solve these problems. By applying design software such as CATIA, Matlab and Simulink, and MSC Adams, a suite of optimization design methods for ECFCRS drive mechanisms are developed here. Based on these new methods, design requirements can be analyzed comprehensively and the design scheme can be modified easily, thus greatly shortening the design cycle. The bench tests and real vehicle tests indicate that the system developed achieves preferable engine speed following-up performance and engine speed regulating performance. The method developed has significance as a reference for developing other vehicle systems.

  14. Sliding Wear and Friction Behavior of Fuel Rod Material in Water and Dry State

    Science.gov (United States)

    Park, Jin Moo; Kim, Jae Hoon; Jeon, Kyeong Lak; Park, Jun Kyu

    In water cooled reactors, the friction between spacer grid and fuel rod can lead to severe wear and it is an important topic to study. In the present study, sliding wear behavior of zirconium alloy was investigated in water and dry state using the pin-on-disc sliding wear tester. Sliding wear resistance of zirconium alloy against heat treated inconel alloy was examined at room temperature. The parameters in this study were sliding velocity, axial load and sliding distance. The wear characteristics of zirconium alloy was evaluated by friction coefficient, specific wear rate and wear volume. The micro-mechanisms responsible for wear in zirconium alloy were identified to be micro-cutting, micro-pitting, delamination and micro-cracking of deformed surface zone.

  15. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  16. Tomography on nuclear fuel rods in the nuclear power plant of Dodewaard. Tomografie aan splijtstofstaven in de centrale Dodewaard

    Energy Technology Data Exchange (ETDEWEB)

    Tanke, R.H.J.; Jaspers, J.E.; Gaalman, P.A.M. (KEMA, Arnhem (Netherlands). Division Research and Development)

    1990-09-06

    This report discusses the feasibility of using emission tomography on fuel rods in the Dodewaard reactor. The tomography can be used to increase the efficiency of the use of fissionable material. (R.A.B.). 4 refs.; 17 figs.; 1 tab.

  17. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs.

  18. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Directory of Open Access Journals (Sweden)

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  19. Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

    Directory of Open Access Journals (Sweden)

    Armando C. Marino

    2011-01-01

    Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

  20. Development of FUELSIM/MOD0 for the detailed analysis of LWR fuel rod behavior under normal operation conditions with extended burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G.A.; Allison, C.M. [Innovative Systems Software LLC, 1284 South Woodruff, Idaho Falls, ID (United States)

    1999-07-01

    The FUELSIM code is being developed by Innovative Systems Software as part of the international SCDAP Development and Training Program. FUELSIM is being developed as a 'stand-alone' best-estimate fuel behavior code with evaluation modeling options. The long term goal of the code is to predict fuel performance over the full range of conditions from normal operating behavior to severe accident conditions using a combination of models from the FRAPCON-3, FRAP-T6, SCDAP, and MATPRO fuel behavior codes. FUELSIM/MOD0 is the first release of the code and includes models to predict the behavior of LWR fuel rods during normal operating conditions including the influence of extended burnup fuel. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The code models all the important phenomena that occur during normal operating conditions and contains necessary materials properties, water properties, and heat transfer correlations. The code runs on a variety of computers and operating systems including UNIX, LINUX, and Windows NT or 95. (author)

  1. Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, Saptarshi, E-mail: saptarshi.bhattacharjee@outlook.com [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)

    2017-06-15

    Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

  2. Effects of gap size and excitation frequency on the vibrational behavior and wear rate of fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Zupan [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Thouless, M.D., E-mail: thouless@umich.edu [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Department of Materials Science & Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Lu, Wei, E-mail: weilu@umich.edu [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States)

    2016-11-15

    Graphical abstract: A wear map shows wear rate as a function of the grid-to-rod gap size and the frequency of the excitation force. The critical gap size, which is associated with the maximum wear rate, lies within the harmonic regime. In the no wear region the amplitude of the rod vibration is smaller than the gap size so that no impact between the rod and plate can happen. The curve of the resonant frequency of the system appears to overlap with the peaks in the contour. - Highlights: • A 3D finite-element based approach to study grid-to-rod fretting. • Two important factors: grid-to-rod gap size and frequency of the excitation force. • Rod vibration shows three regimes: harmonic, period-doubling and chaotic. • A critical gap size is associated with the maximum wear rate. • A wear map shows wear rate as a function of the gap size and excitation frequency. - Abstract: Grid-to-rod fretting (GTRF) wear is a major cause of fuel leaks. Understanding its mechanism is crucial for improving the reliability of nuclear reactors. In this paper we present a three-dimensional, finite-element based approach, which reveals how the wear rate depends on the size of the gap between the grid and the fuel rod, and on the frequency of the excitation force. We show that these two factors affect the dynamic vibration of the rod, which leads to three different regimes: harmonic, period-doubling and chaotic. The wear rate in the harmonic regime is significantly larger than that in the other two regimes, and reaches a maximum when the excitation frequency is close to the resonant frequency of the system, which is dependent on the gap size. We introduce the concept of a critical gap size that gives the maximum wear rate, and we identify the properties and values of this critical gap size. A wear map is developed as a result of a large number of parametric studies. This map shows quantitatively the wear rate as a function of the gap size and excitation frequency, and will be a

  3. Technical Development of Gamma Scanning for Irradiated Fuel Rod after Upgrade of System in Hot-cell

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Seog; Kim, Hee Moon; Baik, Seung Je; Yoo, Byung Ok; Choo, Yong Sun

    2007-06-15

    Non-destructive test system was installed at hot-cell(M1) in IMEF(Irradiated Materials Examination Facility) more than 10 years ago for the diametric measurement and gamma scanning of fuel rod. But this system must be needed to be remodeled for the effective operations. In 2006, the system was upgraded for 3 months. The collimator bench can be movable with horizontal direction(x-direction) by motorized system for sectional gamma scanning and 3-dimensional tomography of fuel rod. So, gamma scanning for fuel rod can be detectable by x, y and rotation directions. It may be possible to obtain the radioactivities with radial and axial directions of pellet. This system is good for the series experiments with several positions. Operation of fuel bench and gamma detection program were linked each other by new program tools. It can control detection and bench moving automatically when gamma inspection of fuel rod is carried out with axial or radial positions. Some of electronic parts were added in PLC panel, and operating panel was re-designed for the remote control. To operate the fuel bench by computer, AD converter and some I/O cards were installed in computer. All of software were developed in Windows-XP system instead of DOS system. Control programs were made by visual-C language. After upgrade of system, DUPIC fuel which was irradiated in HANARO research reactor was detected by gamma scanning. The results were good and operation of gamma scanning showed reduced inspection time and easy control of data on series of detection with axial positions. With consideration of ECT(Eddy Current Test) installation, the computer program and hardware were set up as well. But ECT is not installed yet, so we have to check abnormal situation of program and hardware system. It is planned to install ECT in 2007.

  4. Potential of thorium-based fuel cycle for PWR core to reduce plutonium and long-term toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    The cross section libraries and calculation methods of the participants were inter-compared through the first stage benchmark calculation. The multiplication factor of unit cell benchmark are in good agreement, but there is significant discrepancies of 2.3 to 3.5 %k at BOC and at EOC between the calculated infinite multiplication factors of each participants for the assembly benchmark. Our results with HELIOS show a reasonable agreement with the others except the MTC value at EOC. To verify the potential of the thorium-based fuel to consume the plutonium and to reduce the radioactivity from the spent fuel, the conceptual core with ThO{sub 2}-PuO{sub 2} or MOX fuel were constructed. The composition and quantity of plutonium isotopes and the radioactivity level of spent fuel for conceptual cores were analyzed, and the neutronic characteristics of conceptual cores were also calculated. The nuclear characteristics for ThO{sub 2}-PuO{sub 2} thorium fueled core was similar to MOX fueled core, mainly due to the same seed fuel material, plutonium. For the capability of plutonium consumption, ThO{sub 2}-PuO{sub 2} thorium fuel can consume plutonium 2.1-2.4 times MOX fuel. The fraction of fissile plutonium in the spent ThO{sub 2}-PuO{sub 2} thorium fuel is more favorable in view of plutonium consumption and non-proliferation than MOX fuel. The radioactivity of spent ThO{sub 2}-PuO{sub 2} thorium and MOX fuel batches were calculated. Since plutonium isotopes are dominant for the long-term radioactivity, ThO{sub 2}-PuO{sub 2} thorium has almost the same level of radioactivity as in MOX fuel for a long-term perspective. (author). 22 figs., 11 tabs.

  5. Experimental study of water flow in nuclear fuel elements; Estudo experimental do escoamento de agua em elementos combustiveis nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Lorena Escriche, E-mail: ler@cdtn.br [Centro Federal de Educacao Tecnologica de Minas Gerais (CEFET), Belo Horizonte, MG (Brazil); Rezende, Hugo Cesar; Mattos, Joao Roberto Loureiro de; Barros Filho, Jose Afonso; Santos, Andre Augusto Campagnole dos, E-mail: hcr@cdtn.br, E-mail: jrmattos@cdtn.br, E-mail: jabf@cdtn.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    This work aims to develop an experimental methodology for investigating the water flow through rod bundles after spacer grids of nuclear fuel elements of PWR type reactors. Speed profiles, with the device LDV (Laser Doppler Velocimetry), and the pressure drop between two sockets located before and after the spacer grid, using pressure transducers were measured.

  6. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  7. Fuel utilization improvement in PWRs using the denatured /sup 233/U-Th cycle

    Energy Technology Data Exchange (ETDEWEB)

    Jones, H.M.; Schwenk, G.A.; Toops, E.C.; Yotinen, V.O.

    1980-06-01

    A number of changes in PWR core design and/or operating strategy were evaluated to assess the fuel utilization improvement achievable by their implementation in a PWR using thorium-based fuel and operating in a recycle mode. The reference PWR for this study was identical to the B and W Standard Plant except that the fuel pellets were of denatured (/sup 233/U//sup 238/U-Th)O/sub 2/. An initial scoping study identified the three most promising improvement concepts as (1) a very tight lattice, (2) thorium blankets, and (3) ThO/sub 2/ rods placed in available guide tubes. A conceptual core design incorporating these changes was then developed, and the fuel utilization of this modified design was compared with that of the reference case.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-09

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

  9. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities.

  10. ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Kovscek, S.E.; Martin, S.E.

    1982-10-01

    ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given.

  11. Fuel Performance Characterisation under Various PWR Conditions: Description of the Annealing Test Facilities available at the LECA-STAR laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Pontillon, Y.; Cornu, B.; Clement, S.; Ferroud-Plattet, M.P.; Malgouyres, P.P. [Commissariat a l' Energie Atomique, CEA/DEN/DEC/SA3C - Centre d' Etudes de Cadarache, BP1, 13108 Saint Paul Lez Durance (France)

    2008-07-01

    The aim to improve LWR fuel behaviour has led Cea to improve its post-irradiation examination capacities in term of test facilities and characterization techniques in the shielded hot cells of the LECA-STAR facility, located in Cadarache Cea center. as far as the annealing test facilities are concerned, fuel qualification and improvement of knowledge require a set of furnaces which are already used or will be used. The main characteristics of these furnaces strongly depend on the experimental objectives. The aim of this paper is to review the main aspects of these specific experiments concerning: (i) fission gas release from high burn up fuel, (ii) global fission product release in severe-accident conditions and (iii) fuel microstructural changes, potential cladding failure, radionuclide source terms... under conditions representative of long term dry storage and geological disposal. (authors)

  12. External Attachment of Titanium Sheathed Thermocouples to Zirconium Nuclear Fuel Rods For The Loss-Of-Fluid-Test (LOFT) Reactor

    Science.gov (United States)

    Welty, Richard K.

    1980-10-01

    The Exxon Nuclear Company, Inc. acting as a Subcontractor to EG&G Idaho Inc.3 Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant-accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). The design goals were to (1) reliably attach thermocouples to the zircaloy fuel rods, (2) achieve or exceed a life expectancy of 6,000 hours of reactor operation in a borated water environment of 316°C at 2260 psi, (3) provide and sustain repeatable physical and metallurgical properties in the instrumented rods subjected to transient temperatures up to 1538°C with blowdown, shock, loading, and fast quench. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests.

  13. Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Alexis Weir; Valeria L. Putnam; John D. Bess

    2009-09-01

    The Power Burst Facility (PBF) reactor operated from 1972 to 1985 on the SPERT Area I of the Idaho National Laboratory, then known as Nuclear Reactor Test Station. PBF was designed to provide experimental data to aid in defining thresholds for and modes of failure under postulated accident conditions. PBF reactor startup testing began in 1972. This evaluation focuses on two operational loading tests, chronologically numbered 1 and 2, published in a startup-test report in 1974 [1]. Data for these tests was used by one of the authors to validate a MCNP model for criticality safety purposes [2]. Although specific references to original documents are kept in the text, all the reactor parameters and test specific data presented here was adapted from that report. The tests were performed with operational fuel loadings, a stainless steel in-pile tube (IPT) mockup, a neutron source, four pulse chambers, two fission chambers, and one ion chamber. The reactor's four transition rods (TRs) and control rods (CRs) were present but TR boron was completely withdrawn below the core and CR boron was partially withdrawn above the core. Test configurations differ primarily in the number of shim rods, and consequently the number of fuel rods included in the core. The critical condition was approached by incrementally and uniformly withdrawing CR boron from the core. Based on the analysis of the experimental data and numerical calculations, both experiments are considered acceptable as criticality safety benchmarks.

  14. Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Alexis Weir; Valeria L. Putnam; John D. Bess

    2009-09-01

    The Power Burst Facility (PBF) reactor operated from 1972 to 1985 on the SPERT Area I of the Idaho National Laboratory, then known as Nuclear Reactor Test Station. PBF was designed to provide experimental data to aid in defining thresholds for and modes of failure under postulated accident conditions. PBF reactor startup testing began in 1972. This evaluation focuses on two operational loading tests, chronologically numbered 1 and 2, published in a startup-test report in 1974 [1]. Data for these tests was used by one of the authors to validate a MCNP model for criticality safety purposes [2]. Although specific references to original documents are kept in the text, all the reactor parameters and test specific data presented here was adapted from that report. The tests were performed with operational fuel loadings, a stainless steel in-pile tube (IPT) mockup, a neutron source, four pulse chambers, two fission chambers, and one ion chamber. The reactor's four transition rods (TRs) and control rods (CRs) were present but TR boron was completely withdrawn below the core and CR boron was partially withdrawn above the core. Test configurations differ primarily in the number of shim rods, and consequently the number of fuel rods included in the core. The critical condition was approached by incrementally and uniformly withdrawing CR boron from the core. Based on the analysis of the experimental data and numerical calculations, both experiments are considered acceptable as criticality safety benchmarks.

  15. 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.

  16. CFD modelling of supercritical water flow and heat transfer in a 2 × 2 fuel rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Podila, Krishna, E-mail: krishna.podila@cnl.ca; Rao, Yanfei, E-mail: yanfei.rao@cnl.ca

    2016-05-15

    Highlights: • Bare and wire wrapped 2 × 2 fuel rod bundles were modelled with CFD. • Sensitivity of predictions to SST k–ω, v{sup 2}–f and turbulent Prandtl number was tested. • CFD predictions were assessed with experimentally reported fuel wall temperatures. - Abstract: In the present assessment of the CFD code, two heat transfer experiments using water at supercritical pressures were selected: a 2 × 2 rod bare bundle; and a 2 × 2 rod wire-wrapped bundle. A systematic 3D CFD study of the fluid flow and heat transfer at supercritical pressures for the rod bundle geometries was performed with the key parameter being the fuel rod wall temperature. The sensitivity of the prediction to the steady RANS turbulence models of SST k–ω, v{sup 2}–f and turbulent Prandtl number (Pr{sub t}) was tested to ensure the reliability of the predicted wall temperature obtained for the current analysis. Using the appropriate turbulence model based on the sensitivity analysis, the mesh refinement, or the grid convergence, was performed for the two geometries. Following the above sensitivity analyses and mesh refinements, the recommended CFD model was then assessed against the measurements from the two experiments. It was found that the CFD model adopted in the current work was able to qualitatively capture the trends reported by the experiments but the degree of temperature rise along the heated length was underpredicted. Moreover, the applicability of turbulence models varied case-by-case and the performance evaluation of the turbulence models was primarily based on its ability to predict the experimentally reported fuel wall temperatures. Of the two turbulence models tested, the SST k–ω was found to be better at capturing the measurements at pseudo-critical and supercritical test conditions, whereas the v{sup 2}–f performed better at sub-critical test conditions. Along with the appropriate turbulence model, CFD results were found to be particularly sensitive to

  17. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    Science.gov (United States)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  18. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  19. Experiment data report IFA-226 postirradiation examination. [PWR, BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bagger, C.; Carlsen, H.; Domanus, J.; Hougaard, H.; Larsen, E.; Larsen, N.

    1977-09-01

    IFA-226 contained twelve, mixed plutonium-uranium oxide fuel rods arranged in two, six-rod clusters. The assembly was designed to study fuel-cladding mechanical interaction, fuel thermal response, and fission gas release as a function of fuel density, initial fuel-to-cladding gap, rod power, and burnup. Data were obtained from fuel rod centerline thermocouples, fission gas pressure transducers, and cladding elongation sensors. Results of both nondestructive and destructive examinations are presented. The PIE indicated that one fuel rod failed during service as a result of internal hydriding of the end plug. Circumferential cladding ridges resulting from fuel-cladding interaction were present on all of the rods, with the largest ridges present on the rod with the smallest initial fuel-to-cladding gap. No incipient fuel rod failures were detected.

  20. 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.

  1. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  2. Space and Time Distribution of Pu Isotopes inside The First Experimental Fuel Pin Designed for PWR and Manufactured in Indonesia

    Science.gov (United States)

    Suwardi; Setiawan, J.; Susilo, J.

    2017-01-01

    The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared and planned to be tested in power ramp irradiation. An irradiation test should be designed to allow an experiment can be performed safely and giving maximum results of many performance aspects of design and manufacturing. Performance analysis to the fuel specimen shows that the specimen is not match to be used for power ramp testing. Enlargement by 0.20 mm of pellet diameter has been proposed. The present work is evaluation of modified design for important aspect of isotopic Pu distribution during irradiation test, because generated Pu isotopes in natural UO2 fuel, contribute more power relative to the contribution by enriched UO2 fuel. The axial profile of neutrons flux have been chosen from both experimental measurement and model calculation. The parameters of ramp power has been obtained from statistical experiment data. A simplified and typical base-load commercial PHWR profile of LHR history has been chosen, to determine the minimum irradiation time before ramp test can be performed. The data design and Mat pro XI materials properties models have been chosen. The axial profile of neutrons flux has been accommodated by 5 slices of discrete pin. The Pu distribution of slice-4 with highest power rate has been chosen to be evaluated. The radial discretion of pellet and cladding and numerical parameter have been used the default best practice of TU. The results shows that Pu 239 increased rapidly. The maximum burn up of slice 4 at upper the median slice, it reached nearly 90% of maximum value at about 6000 h with peak of 0.8%a Pu/HM at 22000 h, which is higher than initial U 235. Each 240, 241 and 240 Pu grows slower and ends up to 0.4, 0.2 and 0.18 % respectively. This results can be used for verification of other aspect of fuel behavior in the modeling results and also can be used as guide and comparison to the future post irradiation examination for Pu isotopes distribution.

  3. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  4. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  5. Fuel Management Study on PWR Core Included of 157 Fuel Assemblies%157组燃料组件组成的堆芯燃料管理研究

    Institute of Scientific and Technical Information of China (English)

    姚红

    2013-01-01

    The fuel management of the PWR reactor core reload optimization was studied with SCIENCE codes in the paper ,the PWR core consists of 157 fuel assemblies .The paper studied five strategies ,three strategies are one-year reload and the other two are 18-month reload strategies .The main results of the eight cycles for the five strategies were given ,and the results were compared with each other .In conclusion ,the power peak of the OU T-IN strategy loading pattern is lower ,and the power peak of the IN-OU T loading pattern is higher ,but all of them are satisfied with design limitation .The average discharge burnup of the quarter core strategy is the highest ,which means that the assemblies of this strategy are burned the most sufficiently , so the economic efficiency of the quarter core strategy is the best .%本文应用SCIENCE程序包对157组燃料组件组成的压水堆堆芯进行换料优化燃料管理研究,给出了3个年换料和2个18个月换料共5个设计方案,每个设计方案给出了从首循环到第8循环共8个循环的主要计算结果,并进行了分析比较。综合来看,OUT-IN装载的设计方案功率峰值偏低,IN-OUT装载的设计方案功率峰值偏高,但均在设计限值以内;1/4堆芯换料设计方案的平均卸料燃耗最深,表明其组件燃耗得最充分,经济性较好。

  6. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Ok; Hwang, Mi Jin; Kim, Sung Woo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples.

  7. A study on the direct use of spent PWR fuel in CANDU reactors -Development of DUPIC fuel on manufacturing and quality control technology-

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, Hyun Soo; Lee, Yung Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Oxidation/reduction process was established after analysis of the effect of process parameter on the sintering behavior using SIMFUEL. Process equipment was studied more detail and some of process equipment items were designed and procured. The chemical analysing method of fission products and fissile content in DUPIC fuel was studied and the behavior and the characteristics of fission products in fuel was also done. Requirement for irradiation in HANARO was analysed to prepare performance evaluation. 100 figs, 48 tabs, 170 refs. (Author).

  8. A study on the direct use of spent PWR fuel in CANDU reactors -Development of DUPIC fuel on manufacturing and quality control technology-

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, Hyun Soo; Lee, Yung Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Oxidation/reduction process was established after analysis of the effect of process parameter on the sintering behavior using SIMFUEL. Process equipment was studied more detail and some of process equipment items were designed and procured. The chemical analysing method of fission products and fissile content in DUPIC fuel was studied and the behavior and the characteristics of fission products in fuel was also done. Requirement for irradiation in HANARO was analysed to prepare performance evaluation. 100 figs, 48 tabs, 170 refs. (Author).

  9. Failure behavior of plutonium-uranium mixed oxide fuel under reactivity-initiated accident condition

    Science.gov (United States)

    Abe, T.; Nakae, N.; Kodato, K.; Matsumoto, M.; Inabe, T.

    1992-06-01

    Two series of in-pile tests on MOX fuels were performed in the NSRR to study failure behavior under RIA (reactivity-initiated accident) conditions in water cooled reactors. PWR type MOX test rods were pulsed in a first series. The test rods were designed to have dimensions identical to standard UO 2 fuel, on which a large number of tests had been conducted previously. The test result was that the failure mechanism and the threshold of MOX fuel was consistent with those of UO 2 fuel. ATR-type MOX test rods with PuO 2 particles as well as reference rods without PuO 2 particles were subjected to pulsing in a second series. PuO 2 particles of 400 and 1100 μm in diameter were artificially embedded at the surface of MOX pellets. No effect of particles appeared on the threshold, and no significant indication of their effect was observed on the cladding.

  10. Fluid structure interaction between rods and a cross flow - Numerical approach

    Energy Technology Data Exchange (ETDEWEB)

    Simoneau, Jan-patrice, E-mail: jan-patrice.simoneau@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Sageaux, Thomas, E-mail: thomas.sageaux@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Moussallam, Nadim, E-mail: nadim.moussallam@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Bernard, Olivier, E-mail: olivier.bernard1@areva.com [Areva, 1, Place J. Millet, F 92084 Paris la Defense (France)

    2011-11-15

    This paper presents a full coupled approach between fluid dynamics and structure analysis. It is conducted in order to further improve the assessment of fluid structure interaction problems, occurring in the nuclear field such as the behavior of PWR fuel rods, steam generators and other heat exchangers tubes, fast breeder fuel assemblies. The coupling is obtained by implementing a beam mechanical model in user routines of the CFD code Star-CD, and thanks to a moving grid procedure. The configurations considered are rods in a cross flow. The model is first validated on a single rod case. The lock-in effect is pointed out and both amplitude and frequency responses of the single rod are positively compared to experimental data. Secondly, the mutual influence of two rods, either in-line or parallely set, is investigated. Different behaviors, bounded by critical distances between the rods are highlighted. Finally, the stability of a 3 Multiplication-Sign 3 bundle is calculated for different impinging velocities. Stable and unstable areas are found when varying the impinging velocity. Above a limit, the vibrations amplify up to a contact between rods, this bound is found slightly greater than literature values for close configurations. It is therefore expected that further calculations, with model refinements, will bring valuable informations about bundle stability.

  11. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  12. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  13. Thermomechanical analysis of fuel rods during transitory events using the RAMONA and FETMA codes; Analisis termomecanico de barras combustibles durante eventos transitorios usando los codigos RAMONA y FETMA

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: hector.hernandez@inin.gob.mx

    2009-10-15

    In National Institute of Nuclear Research, the fuel management system (FMS) has been used by long time to simulate the BWR operation in stationary state, as well as during a transitory event. To evaluate the thermomechanical behavior of a fuel element was created and interface between the FMS codes and the fuel element thermo mechanical analysis (FETMA) code properly developed and implemented. In this work, the results of thermomechanical behavior of fuel rods that compose the hot channel during the simulation of a transitory event of a BWR are shown. The transitory events considered in this work are a load rejection and failure in controller of feed water, which are events more important that can to occur in a BWR. The results show that during the developed conditions by both transitory events some failure is not presented in fuel rods. Also, that the transitory event of load rejection is more claimant in security terms that of controller failure of feed water. (Author)

  14. Non-destructive methods of control of thermo-physical properties of fuel rods

    Science.gov (United States)

    Kruglov, A. B.; Kruglov, V. B.; Kharitonov, V. S.; Struchalin, P. G.; Galkin, A. G.

    2017-01-01

    Information about the change of thermal properties of the fuel elements needed for a successful and safe operation of the nuclear power plant. At present, the existing amount of information on the fuel thermal conductivity change and “fuel-shell” thermal resistance is insufficient. Also, there is no technique that would allow for the measurement of these properties on the non-destructive way of irradiated fuel elements. We propose a method of measuring the thermal conductivity of the fuel in the fuel element and the contact thermal resistance between the fuel and the shell without damaging the integrity of the fuel element, which is based on laser flash method. The description of the experimental setup, implementing methodology, experiments scheme. The results of test experiments on mock-ups of the fuel elements and their comparison with reference data, as well as the results of numerical modeling of thermal processes that occur during the measurement. Displaying harmonization of numerical calculation with the experimental thermograms layout shell portions of the fuel cell, confirming the correctness of the calculation model.

  15. A preliminary approach to the extension of the Transuranus code to the fuel rod performance analysis of HLM-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Luzzi, L.; Botazzoli, P.; Devita, M.; Di Marcello, V.; Pastore, G. [Department of Energy, Politecnico di Milano, Enrico Fermi Center for Nuclear Studies - CeSNEF, via Ponzio 34/3, 20133 Milano (Italy)

    2010-07-01

    This paper briefly presents a preliminary modelling approach, aimed at the extension of the TRANSURANUS code to the fuel rod performance analysis of Heavy Liquid Metal (HLM) cooled nuclear reactors, with specific reference to the employment of the T91 steel as cladding material and of the liquid Lead-Bismuth Eutectic (LBE) as coolant. On the basis of literature indications, correlations for heat transfer to LBE, corrosion behaviour and thermo-mechanical properties of T91 are proposed, and some open issues are discussed in prospect of more reliable fuel rod performance analysis of HLM-cooled nuclear reactors. (authors)

  16. Thermoacoustic enhancements for nuclear fuel rods and other high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, Steven L.; Smith, James A.; Kotter, Dale K.

    2017-05-09

    A nuclear thermoacoustic device includes a housing defining an interior chamber and a portion of nuclear fuel disposed in the interior chamber. A stack is disposed in the interior chamber and has a hot end and a cold end. The stack is spaced from the portion of nuclear fuel with the hot end directed toward the portion of nuclear fuel. The stack and portion of nuclear fuel are positioned such that an acoustic standing wave is produced in the interior chamber. A frequency of the acoustic standing wave depends on a temperature in the interior chamber.

  17. Turbulet flow in a model nuclear fuel rod bundle containing partial flow blockages

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1977-03-01

    Local velocity and turbulence intensity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 x 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create area reductions of 70 and 90 percent in the center four subchannels of the bundle. Experimental results indicated that severe flow disturbances existed downstream from the blockage clusters and showed that only minor disturbances can be expected upstream from the blockages. Recirculation zones for both 70 and 90 percent blockages were detected downstream from the blockage clusters and persisted for approximately three to five subchannel hydraulic diameters depending on blockage severity. The experimental velocity results obtained with blockage clusters located midway between grid spacers were successfully predicted using the COBRA computer program.

  18. Effect of Flow Blockage on the Coolability during Reflood in a 2 × 2 Rod Bundle

    Directory of Open Access Journals (Sweden)

    Kihwan Kim

    2014-01-01

    Full Text Available During the reflood phase of a large-break loss-of-coolant accident (LBLOCA in a pressurized-water reactor (PWR, the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

  19. CANDU堆应用RU的PWR/CANDU联合核燃料循环的研究%Study of RU Utilization in CANDU Reactor-an Advanced Nuclear Fuel Cycle of PWR/CANDU Synergism

    Institute of Scientific and Technical Information of China (English)

    霍小东; 谢仲生

    2003-01-01

    对压水堆乏燃料后处理回收铀(RU)在秦山三期CANDU堆中应用的可行性和经济性进行分析.使用ORIGEN2程序,对后处理回收铀在生产后放置不同时间后核素的成份和放射性活度进行了计算.证明RU燃料元件生产的放射性水平是可以接受的.使用DRAGON/DONJON程序对应用RU的秦山三期CANDU堆的时均堆芯和瞬时堆芯校验分析表明:采用简单的2燃耗区,2、4棒束的换料方案能满足最大通道功率、最大棒束功率限制.通过放射性分析和堆芯物理分析可以看出,秦山三期CANDU堆在不改变堆芯结构及运行模式的条件下,从天然铀(NU)燃料过渡到RU燃料是可行的.通过对秦山三期CANDU堆应用RU的经济性分析,可以看出PWR/CANDU联合核燃料循环的策略既可节约铀资源(23%),提高燃料的能量输出(41%),又减少了废燃料的处置量(66%),可大大降低核电成本.

  20. 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.

  1. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  2. Automatic system of welding for nuclear fuel rods; Sistema automatico de soldadura para barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Romero G, M; Romero C, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The welding process of nuclear fuel must be realized in an inert gas environment (He) and constant flow of this. In order to reach these conditions it is necessary to do vacuum at the chamber and after it is pressurized with the noble gas (purge) twice in the welding chamber. The purge eliminates impurities that can provoke oxidation in the weld. Once the conditions for initiating the welding are gotten, it is necessary to draw a graph of the flow parameters, pressure, voltage and arc current and to analyse those conditions in which have been carried out the weld. The rod weld must be free of possible pores or cracks which could provoke rod leaks, so reducing the probability of these failures should intervene mechanical and metallurgical factors. Automatizing the process it allows to do reliable welding assuring that conditions have been performed, reaching a high quality welding. Visually it can be observed the welding process by means of a mimic which represents the welding system. There are the parameters acquired such as voltage, current, pressure and flow during the welding arc to be analysed later. (Author)

  3. 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.

  4. 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.

  5. 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.

  6. Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

    1997-08-01

    Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

  7. PIE results on MOX fuel irradiated in MIHAMA Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yamate, K. [Kansai Electric Power Co., Inc., Mihama, Fukui (Japan). Mihama Power Station; Abeta, S. [Mitsubishi Heavy Industries Ltd. (Japan); Kosaka, Y. [Nuclear Development Corp., Tokai, Ibaraki (Japan); Abe, Y. [Japan Atomic Power Co. (Japan); Kuwahara, H. [Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)

    1995-12-31

    This paper describes the results of the post-irradiation examination (PIE) on the MOX fuel rods irradiated in the Japanese commercial PWR, MIHAMA Unit 1. The objective of the PIE is not only to confirm the fuel integrity but also to build up the irradiation data base for MOX fuel design. After three cycles irradiation up to the assembly burnup of 23 GWd/t, they were examined at the site and further PIE was carried out on eight MOX fuel rods at a hot laboratory. The non-destructive test results of PIE proved the integrity of the MOX fuel rods up to the burnup of about 25GWd/t, and revealed the similar irradiation behaviour of dimensional change with the standards UO{sub 2} fuel rods. Results of the following destructive tests also revealed the similar irradiation behaviour of FGR, fuel pellet dimensional change and cladding oxidation with the standard UO{sub 2} fuel, and confirmed no abnormality in microstructure changes of fuel pellets. (author).

  8. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage; Estudo de um casco nacional e sua instalacao para armazenagem seca de combustivel nuclear queimado gerado em reatores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio

    2009-07-01

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  9. Effect of fission fragment on thermal conductivity via electrons with an energy about 0.5 MeV in fuel rod gap

    Directory of Open Access Journals (Sweden)

    F Golian

    2017-02-01

    Full Text Available The heat transfer process from pellet to coolant is one of the important issues in nuclear reactor. In this regard, the fuel to clad gap and its physical and chemical properties are effective factors on heat transfer in nuclear fuel rod discussion. So, the energy distribution function of electrons with an energy about 0.5 MeV in fuel rod gap in Busherhr’s VVER-1000 nuclear reactor was investigated in this paper. Also, the effect of fission fragments such as Krypton, Bromine, Xenon, Rubidium and Cesium on the electron energy distribution function as well as the heat conduction via electrons in the fuel rod gap have been studied. For this purpose, the Fokker- Planck equation governing the stochastic behavior of electrons in absorbing gap element has been applied in order to obtain the energy distribution function of electrons. This equation was solved via Runge-Kutta numerical method. On the other hand, the electron energy distribution function was determined by using Monte Carlo GEANT4 code. It was concluded that these fission fragments have virtually insignificant effect on energy distribution of electrons and therefore, on thermal conductivity via electrons in the fuel to clad gap. It is worth noting that this result is consistent with the results of other experiments. Also, it is shown that electron relaxation in gap leads to decrease in thermal conductivity via electrons

  10. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  11. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  12. Evaluation of alternative treatments for spent fuel rod consolidation wastes and other miscellaneous commercial transuranic wastes

    Energy Technology Data Exchange (ETDEWEB)

    Ross, W.A.; Schneider, K.J.; Oma, K.H.; Smith, R.I.; Bunnell, L.R.

    1986-05-01

    Eight alternative treatments (and four subalternatives) are considered for both existing commercial transuranic wastes and future wastes from spent fuel consolidation. Waste treatment is assumed to occur at a hypothetical central treatment facility (a Monitored Retrieval Storage facility was used as a reference). Disposal in a geologic repository is also assumed. The cost, process characteristics, and waste form characteristics are evaluated for each waste treatment alternative. The evaluation indicates that selection of a high-volume-reduction alternative can save almost $1 billion in life-cycle costs for the management of transuranic and high-activity wastes from 70,000 MTU of spent fuel compared to the reference MRS process. The supercompaction, arc pyrolysis and melting, and maximum volume reduction alternatives are recommended for further consideration; the latter two are recommended for further testing and demonstration.

  13. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    OpenAIRE

    Giovedi Claudia; Cherubini Marco; Abe Alfredo; D’Auria Francesco

    2016-01-01

    Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel) program are considering di...

  14. A comparison of crud phases appearing on some Swedish BWR fuel rods using Laser Raman Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden)]|[Lulea Univ. of Technology (Sweden)

    2002-07-01

    Previous investigations showed that laser Raman spectroscopy (LRS) can be used as a phase specific analytical tool for radioactive fuel crud samples and also for details in the underlying layer of zirconium dioxide. It is relatively easy to record Raman spectra that discriminate between chemical phases for all crud oxides of interest. The method has therefore been recommended for crud investigations within the Swedish program. At ideal conditions the resolution is about 1 {mu}m, permitting detailed position determination of crud phases in the sample. Therefore LRS is a very good complement to X-ray diffraction (XRD). The methods for sample preparation and handling of radioactive crud samples for LRS turn out to be relatively simple. A detailed LRS study on fuel crud samples from Barsebaeck 2, Forsmark 2, Forsmark 3 and Ringhals 1 was performed in this work. All of those Swedish BWRs were operated at different conditions at the time of sampling. The chemistry regimes covered NWC, HWC and other variable conditions. Also different types of fuel, exposure times and sampling positions were selected. (authors)

  15. FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

    Energy Technology Data Exchange (ETDEWEB)

    Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Shestopalov, A., E-mail: shest@dhtp.kiae.ru [RRC' Kurchatov Institute' , Kurchatov sq, 123182 Moscow (Russian Federation)

    2011-07-15

    Highlights: > We developed a new fuel rod behaviour code named FRED. > Main models and assumptions are described. > The code was checked using the IFA-503.2 tests performed at the Halden reactor. - Abstract: The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.

  16. The Recovery of the Metal Insulation Cable in the Instrumentation of Nuclear Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Sim, Bong Sik; Lee, Chul Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Mineral-insulated (MI) cables are widely used to prolong the instrumentation cable of instruments such as a thermocouple (TC), linear variable differential transformer (LVDT) and self-powered neutron detector (SPND), which are used to measure various irradiation characteristics of nuclear fuels and materials. MI cables are expected to be helpful for instrumentation of nuclear fuel and material irradiation because of their high electrical insulation, heat resistance and mechanical strength. The MI cable used to extend thermocouple wires is classified as the following: 1) For common metal types of thermocouples, the thermocouple extension wire is of substantially the same composition as the corresponding thermocouple type and it can offer advantages in cost or mechanical properties when used for the connection between a thermocouple and instruments. 2) For noble metal types of thermocouples, the thermocouple compensation wire is an entirely different alloy formulated to match the noble metal characteristics, which is necessary due to the high cost of noble metals. During the installation of an instrument, an MI cable damaged by impact must be recovered because it is difficult to change the entire thermocouple. And for MI cable recovery, it is necessary to develop the instrumentation technology of FTL. This paper described the experimental results of MI cable recovery, which consists of a removal test of the MI cable sheath and a joining test of the compensation of the wire and MI cable/ wire/compensation wire and sheath of MI cable/bushing, for carrying out irradiation tests of nuclear fuel and materials in the FTL facility of HANARO

  17. Anisotropic Azimuthal Power and Temperature distribution on FuelRod. Impact on Hydride Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Motta, Arthur [Pennsylvania State Univ., State College, PA (United States); Ivanov, Kostadin [Pennsylvania State Univ., State College, PA (United States); Arramova, Maria [Pennsylvania State Univ., State College, PA (United States); Hales, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-29

    The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split into two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.

  18. Investigation of the structure of debris beds formed from fuel rods fragmentation

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Duc-Hanh; Fichot, Florian; Topin, Vincent, E-mail: vincent.topin@irsn.fr

    2017-03-15

    This paper is a study of debris beds that can form in the core of a nuclear power plant under severe accident conditions. Such beds are formed of fragments of pellets and cladding remnants, as observed in the TMI-2 core. Many important issues are related with the morphology of those debris beds: are they coolable in case of water injection and how does molten corium progress through them if they are not coolable? The answers to those questions depend on the structure of the debris bed: porosity, number and arrangement of particles. In order to obtain relevant information, a numerical simulation of the formation of the debris bed is proposed. It relies on a granular approach of the type called “Contact Dynamics” to simulate the collapse of debris and their accumulation. Two different schemes of fuel pellet fragmentation are considered and simulations for different degrees of fragmentation of the pellets are performed. The results show that the number of axial cracks on fuel pellets strongly influences the final porosity of the debris bed. Porosities vary between 31% (less coolable cases) and 45% (similar to TMI-2 observations), with a most probable configuration around 41%. The specific surface of the bed is also evaluated. In the last part, a simple model is used to estimate the impact of the variation in geometry of the numeric debris beds on their flow properties. We show that the permeability and passability can vary respectively with a range of 30% and 15% depending on the number of fragment per pellet. The other benefits of the approach are finally discussed. Among them, the possibility to print 3D samples from the calculated images of debris beds appears as a promising perspective to perform experiments with realistic debris beds.

  19. 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.

  20. 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.

  1. Disposal Of Spent Fuel In Salt Using Borehole Technology: BSK 3 Concept

    Energy Technology Data Exchange (ETDEWEB)

    Fopp, Stefan; Graf, Reinhold [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany); Filbert, Wolfgang [DBE TECHNOLOGY GmbH, Eschenstrasse 55, D-31224 Peine (Germany)

    2008-07-01

    The BSK 3 concept was developed for the direct disposal of spent fuel in rock salt. It is based on the conditioning of fuel assemblies and inserting fuel rods into a steel canister which can be placed in vertical boreholes. The BSK 3 canister is suitable for spent fuel rods from 3 PWR or 9 BWR fuel assemblies. The emplacement system developed for the handling and disposal of BSK 3 canisters comprises a transfer cask which provides appropriate shielding during the transport and emplacement process, a transport cart, and an emplacement device. Using the emplacement device the transfer cask will be positioned onto the top of the borehole lock. The presentation describes the development and the design of the transfer cask and the borehole lock. A technically feasible and safe design for the transfer cask and the borehole lock was found regarding the existing safety requirements for radiation shielding, heat dissipation and handling procedure. (authors)

  2. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Terttaliisa, E-mail: terttaliisa.lind@psi.c [Paul Scherrer Institut, Villigen (Switzerland); Csordas, Anna Pinter; Nagy, Imre [HAS KFKI Atomic Energy Research Institute, Budapest (Hungary); Stuckert, Juri [Forschungszentrum Karlsruhe, Karlsruhe (Germany)

    2010-02-15

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T approx 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T approx 1650 K, followed by a

  3. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Science.gov (United States)

    Lind, Terttaliisa; Csordás, Anna Pintér; Nagy, Imre; Stuckert, Juri

    2010-02-01

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ˜ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ˜ 1650 K, followed by a relatively

  4. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  5. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2012-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  6. 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

  7. Sensitivity study for accident tolerant fuels: Property comparisons and behavior simulations in a simplified PWR to enable ATF development and design

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Kristina Yancey, E-mail: kristina.yancey@gmail.com; Sudderth, Laura; Brito, Ryan A.; Evans, Jordan A.; Hart, Clifford S.; Hu, Anbang; Jati, Andi; Stern, Karyn; McDeavitt, Sean M., E-mail: mcdeavitt@tamu.edu

    2016-12-01

    Highlights: • This study compared four accident tolerant fuels against uranium dioxide. • Material property correlations were developed to evaluate fuel performance. • The fuels’ neutronic and thermal hydraulic behaviors were studied in the AP1000. • No fuel type performed better in all areas, but each has strengths and weaknesses. • More research is needed to build a complete model of the fuel performances. - Abstract: Since the events at the Fukushima-Daiichi nuclear power plant, there has been increased interest in developing fuels to better withstand accidents for current light water reactors. Four accident tolerant fuel candidates are uranium oxide with beryllium oxide additives, uranium oxide with silicon carbide matrix additives, uranium nitride, and uranium nitride with uranium silicide composite. The first two candidates represent near-term high performance uranium oxide with high thermal conductivity and neutron transparency, and the second two represent mid-term high-density fuels with highly beneficial thermal properties. This study seeks to understand the benefits and drawbacks of each option in place of uranium dioxide. To assess the material properties for each of the fuel types, an extensive literature review was performed for material property data. Correlations were then made to evaluate the properties during reactor operation. Neutronics and thermal hydraulics studies were also completed to determine the impact of the use of each candidate in an AP1000 reactor. In most cases, the candidate fuels performed more desirably than uranium dioxide, but no fuel type performed better in all aspects. Much more research needs to be performed to build a complete model of the fuel performances, primarily experimental data for uranium silicide. Each of the fuels studied has its own benefits and drawbacks, and the comparisons discussed in this report can be used to aid in determining the most appropriate fuel depending on the desired specifications.

  8. 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

  9. 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.

  10. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  11. Thermomechanical analysis of a fuel rod in a BWR reactor using the FUELSIM code; Analisis termomecanico de una barra de combustible de un reactor BWR utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, IPN, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico, D. F. (Mexico); Ortiz V, J.; Araiza M, E. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rapaca78@yahoo.com.mx

    2009-10-15

    The thermomechanical behaviour of a fuel rod exposed to irradiation is a complex process in which are coupled great quantity of interrelated physical-chemical phenomena, for that analysis of rod performance in the core of a nuclear power reactor is realized generally with computation codes that integrate several phenomena expected during the time life of fuel rod in the core. An application of this type of thermomechanical codes is to predict, inside certain reliability margin, the design parameters that would be required to adjust, in order to get a better economy or rod performance, for a systematic approach to the fuel design optimization. FUELSIM is a thermomechanical code based on the models of FRAPCON code, which was developed under auspice of Nuclear Regulatory Commission of USA. FUELSIM allows iterative calculations like part of its programming structure, allowing search of extreme cases of behaviour, probabilistic analysis (or statistical), parametric analysis (or sensibility) and also can include as entrance data to the uncertainties associated with production data, code parameters and associated models. In this work is reported a first analysis of thermomechanical performance of a typical fuel rod used in a BWR 5/6. Results of maximum temperatures are presented in the fuel center and of axial deformation, for the 10 axial nodes in that the active longitude of fuel rod was divided. (Author)

  12. Power Flattening and Rejuvenation of PWR Spent Fuel Blanket for Hybrid Fusion-Fission Reactor%功率展平的压水堆乏燃料发电包层中子学初步研究

    Institute of Scientific and Technical Information of China (English)

    马续波; 陈义学; 王继亮; 王悦; 韩静茹; 陆道纲

    2011-01-01

    The hybrid fusion-fission reactor has advantages of breeding of the nuclear fuel and transmutation of the long-life nuclear waste and having inherent safety. Meanwhile, the engineering and technological demand of hybrid reactor is significantly reduced comparing with that of pure fusion reactor. A generating electricity blanket concept using the PWR spent fuel directly was proposed, which was based on ITER parameter level achieved. Different volume fractions of the fuel in blanket enabled to realize a power flattening in the fissile zone. The results show that the peak-to-average power factor becomes less than no power flattening, and the output power of the fuel zone raises more than 21. 7%. At the end of the operation, the maximum fuel enrichment is 5. 23%. The blanket is feasible from the neutronics viewpoint.%聚变裂变混合堆在增殖核燃料、嬗变长寿命核废料及固有安全性等方面具有较大优势,同时,它比纯聚变堆在工程及技术方面要求低,因此较聚变堆更易实现.本工作基于目前国际聚变实验堆(ITER)所能达到的技术水平,提出一种直接利用乏燃料进行发电的聚变裂变混合堆包层概念,利用在不同位置放置不同乏燃料体积分数的方法对燃料增殖区实现了功率展平.计算结果表明:功率展平后的包层功率不均匀系数更小,且包层中燃料区的能量输出要比不展平情况下的能量输出高约21.7%.燃料富集度到运行末期最大可达5.23%.从中子学角度初步论证了该包层的可行性.

  13. Development of fuel performance and thermal hydraulic technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  14. Development and experimental verification of SST-GRASS: a steady-state and transient fuel response and fission-product release code. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J.; Seitz, M.G.; Gehl, S.M.; Kelman, L.R.

    1976-01-01

    A comprehensive fission-product release model (GRASS), based on a mechanistic understanding of fuel behavior in LWR fuel elements for a wide range of accidental overheating conditions as well as steady-state irradiations, is being developed at Argonne National Laboratory. Experimental support for GRASS is provided by out-of-reactor transient heating of irradiated commercial LWR fuel using a direct-electrical-heating technique. The GRASS calculations are described, benchmarked against standard theoretical treatments, and verified for steady-state irradiations. In addition, preliminary results from the direct-electrical-heating experiments are reported. Possible mechanisms for fission-gas release during transient heating of LWR fuel are discussed based on comparisons of GRASS results with experimental observations.

  15. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  16. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  17. Status of rod consolidation, 1988

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1989-01-01

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs.

  18. Determination of internal pressure and the backfill gas composition of nuclear fuel rods; Determinacion de la presion interna y la composicion del gas de llenado de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, M.A.; Cota S, G.; Merlo S, L.; Fernandez T, F. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    An important consideration in the nuclear fuel manufacturing is the measurement of the helium atmosphere pressure and its composition analysis inside the nuclear fuel rod. In this work it is presented a system used to measure the internal pressure and to determine the backfill gas composition of fuel rods. The system is composed of an expansion chamber provided of a seals system to assure that when rod is drilled, the gas stays contained inside the expansion chamber. The system is connected to a pressure measurement digital system: Baratron MKS 310-AHS-1000. Range 1000 mm Hg from which the pressure readings are taken when this is stabilized in all the system. After a gas sample is sent toward a Perkin Elmer gas chromatograph, model 8410 with thermal conductivity detector to get the corresponding chromatogram and doing the necessary calculations for obtaining the backfill gas composition of the rod in matter. (Author)

  19. Conceptual Design of Structural Components of a Dual Cooled Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Kang-Hee; Kim, Jae-Yong; Yoon, Kyung-Ho

    2008-01-15

    A dual cooled fuel, featured by an internal as well as an external coolant flow passage of a fuel rod, was suggested to enable a large-scaled power-uprate of PWR plant and launched as one of the National Nuclear R and D Projects in 2007. It is necessary to make the dual cooled fuel be compatible with an OPR-1000 system to maximize the economy. Also, the structural components of the dual cooled fuel should be designed to realize their features. To this end, a conceptual design of a spacer grid, outer and center guide tubes, and top and bottom end pieces has been carried out in the project 'Development of Design Technology for Dual Cooled Fuel Structure'. For the spacer grids, it is suggested that springs and dimples are located at or near the cross points of the straps due to a considerably narrowed rod-to-rod gap. Candidate shapes of the grids were also developed and applied for domestic patents. For the outer and center guide tubes, a dual tube like a fuel rod was suggested to make the subchannel areas around the guide tubes be similar to those around the fuel rods of enlarged diameter. It was applied for the domestic patent as well. For the top and bottom end pieces, the shape and pattern have been changed from the conventional ones reflecting the fuel rods' changes. Technical issues and method of resolution for each components were listed up for a basic design works in the following years.

  20. A sensitivity study on DUPIC fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong

    1997-01-01

    In DUPIC fuel cycle, the spent pressurized water reactor (PWR) fuel is refabricated as a DUPIC fuel by a dry process. Because the spent PWR fuel composition depends on the initial enrichment and burnup condition of PWR fuel, the composition of a DUPIC fuel is not uniquely defined. Therefore, for the purpose of reducing the effects of such a composition heterogeneity on core performance, a composition adjustment of DUPIC fuel was studies. The composition adjustment was made in two steps: mixing two spent PWR fuel assemblies of higher and lower {sup 239}Pu contents and blending in fresh uranium with the mixed spent PWR fuels. Because the fuel and core performances depend on both the absolute amount of fissile isotopes and the ratio of major fissile isotope contents, a parametric study was performed to determine the reference compositions of {sup 235}U and {sup 239}Pu. The reference enrichments of {sup 235}U and {sup 239}Pu were determined such that the DUPIC core performance is comparable to that of a natural uranium core with high spent PWR fuel utilization and low fuel cycle cost. Under this condition, it is possible to utilize 90% of spent PWR fuels as the DUPIC fuel formula. On average, the amounts of slightly enriched and depleted uranium used for blending correspond to 8.6% and 10.6%, respectively, of the mass of candidate spent PWR fuels. (author). 16 refs., 30 tabs., 9 figs.

  1. DESCRIPTION OF THE TRITIUM-PRODUCING BURNABLE ABSORBER ROD FOR THE COMMERCIAL LIGHT WATER REACTOR TTQP-1-015 Rev 19

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A.; Love, Edward F.; Thornhill, Cheryl K.

    2012-02-01

    Tritium-producing burnable absorber rods (TPBARs) used in the U.S. Department of Energy’s Tritium Readiness Program are designed to produce tritium when placed in a Westinghouse or Framatome 17x17 fuel assembly and irradiated in a pressurized water reactor (PWR). This document provides an unclassified description of the current design baseline for the TPBARs. This design baseline is currently valid only for Watts Bar reactor production cores. A description of the Lead Use TPBARs will not be covered in the text of the document, but the applicable drawings, specifications and test plan will be included in the appropriate appendices.

  2. Basic research and industrialization of CANDU advanced fuel - Effect of transverse convex curvature on boiling heat transfer and ONB point of nucleate fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Chun; Lee, Young; Lee, Sung Hong [Pusan National University, Pusan (Korea)

    2000-04-01

    Recently, the effect of convex curvature on heat transfer should not be ignored when the radius of curvature tends to be small and/or associated with high heat transfer rate cases. Both analytical and experimental studies were performed to prove the effect of transverse convex curvature on the boiling heat transfer in concentric annuli flows. The effect of the transverse convex surface curvature on ONB are studied analytically in the case of reactor and evaporator. It is seen that the inner wall heat flux depends on R/sub i/, Rc, Re, Pr, {alpha}, and the {theta} of working fluid. An experimental study on the incipience of nucleate boiling is performed as a verification ad extension of previous analyses. Through flow visualization, the results show that the most dominant parameter to affect the heat flux at ONB is found to be the surface curvature. The heat flux data at ONB increases with the Re and the subcooling, and the effect of subcooling on ONB becomes smaller with decreasing Re. The heat flux at ONB increases rapidly as increase in {alpha} due to higher convective motion of bulk flow. Comparison between both results are accomplished with respect to the relative enhancement due to the convex curvature. The relative heat transfer enhancement ratio shows a good agreement between theory and experiment qualitatively and quantitatively. In conclusion, the obtained results suggest that the effect transverse convex curvature appears significantly in the boiling heat transfer. Therefore, it can be clearly expected that the effect should be more strong at the case of critical heat flux condition which is the most important design goal of the advanced nuclear fuel rods. 30 refs., 78 figs. (Author)

  3. 燃料棒径向温度场稳态计算分析%Calculation and Analysis of the Radial Temperature Field of the Fuel Rods

    Institute of Scientific and Technical Information of China (English)

    齐航; 周蓝宇; 张雍良; 曾文杰

    2016-01-01

    燃料棒是反应堆的核心部件,其内部温度场分布大都通过数值计算获得。以燃料棒为研究对象,以燃料棒中心为起点,在径向上划分足够多的环形区域,建立几何模型,依据几何模型建立堆芯稳态物理模型,通过编程进行数值计算来获得燃料元件的径向稳态温度场。以次临界堆MYRRHA的燃料棒为研究对象,研究结果表明该方法能较准确的表征燃料元件径向稳态温度场的情况,是一种简单有效的建模分析方法。可见,该模型可以为燃料元件径向稳态温度场计算提供合理的依据。%Fuel rods is the core component of the reactor, often, its inner temperature field distribution is obtained through numerical calculation method. Taking the fuel rod as the research object, the center of the fuel rod as the starting point, division enough annular region in the radial, and the geometric model is set up, according to the geometric model building reactor core steady-state physical model, apply numerical calculation and programming to obtain fuel element radial steady-state temperature field. Sub-critical reactor MYRRHA fuel element as the research object. The results show that the method can accurately characterize the radial temperature field of the cylindrical fuel element, and it is a simple and effective modeling and analysis method. It can be seen that the model can provide a reasonable basis for calculating the radial temperature field of the cylindrical fuel element.

  4. Light water reactor fuel element suitable for thorium employment in a discrete seed and blanket configuration with the aim to attain conversion ratios above the range of one

    Energy Technology Data Exchange (ETDEWEB)

    Hrovat, M.F.; Grosse, K.H.; Seemann, R. [ALD Vacuum Technologies GmbH, Hanau (Germany)

    2008-07-01

    The thorium resources in the world are relatively large. According to the IAEA-NEA-publication ''Red Book'' they amount to 4.5 10E6 metric tons and are about 4 times greater than the resources of Uranium. The fuel element described in this paper could be used in light water reactor (LWR) preferably in pressurized water reactor (PWR). The seed (feed) rods contain uranium 235 as fissionable material and the blanket (breed) rods contain thorium and uranium. The thorium in the blanket rods is converted to fissionable U-233 by irradiation with thermal neutrons. The U-233 produced is a valuable fissionable material and is characterized by high revalues, where t is defined as the number of fission neutrons per absorption in fissile materials. By optimized configuration and loading of the seed- and blanket rods the thorium is converted to U-233 and the U-238 is converted to fissionable Plutonium isotopes. Consequently more fissionable material is generated than is used. The fuel cycle is also flexible. Thus U-235, Pu-239 or weapons-grade Plutonium can be used.Based on knowledge obtained in the development of fuel elements for material test reactors (MTR), high temperature reactors (HTR) and light water reactors (LWR), a new design of fuel element suitable for thorium employment in PWR is described.

  5. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  6. Analysis on Fuel Thermal Conductivity Model of the Computer Code for Performance Prediction of Fuel Rods%燃料元件性能分析程序中的燃料热导率模型分析

    Institute of Scientific and Technical Information of China (English)

    李海; 黄晨; 杜爱兵; 徐宝玉

    2014-01-01

    The thermal conductivity is one of the most important parameters in the computer code for performance prediction for fuel rods.Several fuel thermal conductivity models used in foreign computer code,including thermal conductivity models for MOX fuel and UO2 fuel were introduced in this paper. Thermal conductivities were calculated by using these models, and the results were compared and analyzed.Finally, the thermal conductivity model for the native computer code for performance prediction for fuel rods in fast reactor was recommended.%热导率是燃料元件性能分析程序最重要的参数之一,本文介绍了各国部分性能分析程序的燃料热导率模型,按照 MOX和 UO2燃料分类,给出了这些性能分析程序热导率模型的计算结果,并进行分析对比,给出了国产快堆性能分析程序的热导率推荐模型。

  7. Study of an ADS Loaded with Thorium and Reprocessed Fuel

    Directory of Open Access Journals (Sweden)

    Graiciany de Paula Barros

    2012-01-01

    Full Text Available Accelerator-driven systems (ADSs are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods was T232hO2 for U233 production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use of T232hO2 without the initial requirement of U233 enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations. The multiplication factor (keff evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use of T232hO2 and reprocessed fuel allowed U233 production without the initial requirement of U233 enrichment.

  8. 水堆燃料元件性能分析及程序FROBA开发%Analysis of Fuel Rod Behavior and Design of FROBA Code

    Institute of Scientific and Technical Information of China (English)

    杨震; 苏光辉; 田文喜; 秋穗正

    2012-01-01

    在详细分析芯块和包壳的辐照行为的基础上,开发了燃料元件性能分析程序FROBA,并对燃料元件的热工-机械-材料特性进行模拟分析,计算得到不同燃耗深度下燃料元件的温度、应变特性.通过与美国爱达荷国家实验室的软件计算结果进行对比,验证本工作开发程序的准确性.结果表明:在芯块和包壳接触前,芯块温度先上升,密实化消失后温度逐渐下降;接触后芯块温度会再次上升.%The temperature and strain profile of pellet and cladding were studied by developing a thermomechanic coupling code FROBA,which was based on analyzing fuel rod behavior theoretically during irradiation. Based on the analysis of results under different operating conditions, a numerical method for calculating fuel rod behavior was obtained, which could be used for the analysis of fuel component under operational conditions of nuclear reactors. The reliability of the code was also proved by comparing the results derived from Idaho National Laboratory software. The results show that the fuel temperature rises before irradiation. Once the densification is complete, the fuel temperature drops. After the gap closure occurs, the temperature gradually rises again.

  9. Understanding the Atomic-Level Chemistry and Structure of Oxide Deposits on Fuel Rods in Light Water Nuclear Reactors Using First Principles Methods

    Science.gov (United States)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.; Andersson, D. A.; Stanek, C. R.

    2016-09-01

    The results of recent studies are discussed in which first principles calculations at the atomic level have been used to expand the thermodynamic database for science-based predictive modeling of the chemistry, composition and structure of unwanted oxides that deposit on the fuel rods in pressurized light water nuclear reactors. Issues discussed include the origin of the particles that make up deposits, the structure and properties of the deposits, and the forms by which boron uptake into the deposits can occur. These first principles approaches have implications for other research areas, such as hydrothermal synthesis and the stability and corrosion resistance of other materials under other extreme conditions.

  10. Control Rod Ejection Accident while Using 6- and 8-Tube IRT-4M Fuel Assemblies in WWR-SM Research Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Baytelesov, S.; Kungurov, F.; Safarov, A.; Salikhbaev, U.

    2011-07-01

    The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel in 2009. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of control rod ejection transient analysis for these mixed cores

  11. Radiative heat transfer modelling in a PWR severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Magali Zabiego; Florian Fichot [Institut de Radioprotection et de Surete Nucleaire - BP 3 - 13115 Saint-paul-Lez-Durance (France); Pablo Rubiolo [Westinghouse Science and Technology - 1344 Beulah Road - Pittsburgh - PA 15235 (United States)

    2005-07-01

    Full text of publication follows: The present study is devoted to the estimation of the radiative heat transfers during a severe accident sequence in a Pressurized Water Reactor. In such a situation, the residual nuclear power released by the fuel rods can not be evacuated and heats up the core. As a result, the cylindrical rods and the structures initially composing the core undergo a degradation process: swelling, breaking or melting of the rods and structures and eventual collapse to form a heap of fragments called a debris bed. As the solid matrix loses its original shape, the core geometry continuously evolves from standing, regularly-spaced cylinders to a non-homogeneous system including deformed remaining rods and structures and debris particles. To predict this type of sequence, the ICARE/CATHARE software [1] is developed by IRSN. Since the temperatures can reach values greater than 3000 K, it was of major interest to provide the code with an accurate radiative transfer model usable whatever the geometry of the system. Considering the size of a reactor core compared to the mean penetration length of radiation, the core can be seen as an optically thick medium. This observation led us to use the diffusion approximation to treat the radiation propagation. In this approach, the radiative flux is calculated in a way similar to thermal conduction: q{sub r} = [K{sub e}].{nabla}T where [K{sub e}] is the equivalent conductivity tensor of the system accounting for thermal and radiative transfer. An homogenization technique is applied to estimate the equivalent conductivity. Given the temperature level, the radiative contribution to the equivalent conductivity tensor quickly becomes dominant. This model was described earlier in [2] in which it was shown that an equivalent conductivity can be continuously calculated in the system when the geometry evolves from standing regular cylinder rods to swollen or broken ones, surrounded or not by a film of liquid materials, to

  12. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  13. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  14. Study on PWR Thorium-uranium Breeding Cycle Using Uniformly Mixed Fuel Assembly%使用均匀混合型燃料组件的压水堆钍-铀增殖循环研究

    Institute of Scientific and Technical Information of China (English)

    周明; 沈季; 于悦海; 张文杰

    2014-01-01

    In order to improve the utilization rate of PWR nuclear fuel , a kind of uniformly mixed thorium-uranium assemblies which contain suitable quantity of 232 Th and 233 U were developed .Neutronics calculation and analysis show that kinf of the new assemblies decreases with the increase of burnup slowly .This property is very good for extending reactor core cycle lifetime .Unit 1 of Ling Ao Nuclear Power Plant was chosen as reference core , and thorium-uranium mixed core was formed by feeding thorium assemblies .Through corresponding analysis ,the conclusion indicates that uniformly mixed thorium-uranium assemblies suit exiting PWRs and have advantages in 235 U utilization and long cycle lifetime was obtained .%为提升压水堆燃料利用率,设计了一种包含适量232 T h和233 U的均匀混合型燃料组件。对该型燃料组件的核特性分析表明,其具备随燃耗增加 kinf下降更缓慢的特性,有利于堆芯获得更长的循环长度。以岭澳核电厂一号机组为例,对包含均匀混合型含钍燃料组件的堆芯进行了分析,结果表明,当前压水堆中采用均匀混合型含钍燃料组件是可行的,并且具备235 U利用率高、堆芯循环长度长的优势。

  15. Realistic bandwidth estimation in the theoretically predicted radionuclide inventory of PWR-UO2 spent fuel derived from reactor design and operating data

    Energy Technology Data Exchange (ETDEWEB)

    Fast, Ivan

    2017-06-01

    Nuclear energy for power generation produces heat-generating high- and intermediate level radioactive waste (HLW and ILW) for which a safe solution for the handling and disposal has to be found. Currently, many European countries consider the final disposal of HLW and ILW in deep geological formations as the most preferable option. In Germany the main stream of HLW and ILW include spent fuel assemblies from nuclear power plants (NPPs), the vitrified waste and compacted metallic waste of the fuel assembly structural parts originate from reprocessing plants. An important task that occurs within the framework of the Product Quality Control (PQC) of nuclear waste is the assessment of the compliance of any reprocessed waste product inventory with the prescribed limits for each relevant radionuclide (RN). The PQC task is to verify the required quality and safety of nuclear waste prior to transportation to a German repository and to avert the disposal of non-conform waste packages. The verification is usually based on comparing the declared radionuclide inventory of the waste with the presumed or expected composition, which is estimated, based on the known history of the waste and its processing. The difficulty of such estimations for radioactive components from nuclear fuel assemblies is that reactor design parameters and operating histories can have a significant influence on the nuclide inventory of any individual fuel assembly. Thus, knowledge of these parameters is a key issue to determine the realistic concentration ranges, or bandwidths, of the radionuclide inventory. As soon as a governmental decision on the construction of a high-level waste repository will be made, comprehensive radionuclide inventories of the wastes assigned for the deposition will be required. The list of final repository relevant radionuclide is based on the safety assessment for this particular repository, thus it is likely to comprise more-or-less the same radionuclides that need to be

  16. Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L

    1998-11-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  17. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  18. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified; Estudo da transferencia de calor em varetas combustiveis 3D do reator EPRI-9R 3D modificado

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: deisedy@gmail.com, E-mail: diogosb@outlook.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results.

  19. AgInCd control rod failure in the QUENCH-13 bundle test

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)], E-mail: leo.sepold@imf.fzk.de; Lind, T. [Paul Scherrer Institut, Laboratory for Thermalhydralics (LTH), Department of Nuclear Energy and Safety (NES), 5232 Villigen PSI (Switzerland); Csordas, A. Pinter [Fuel Materials Department, HAS KFKI AEKI, 1121 Budapest (Hungary); Stegmaier, U.; Steinbrueck, M.; Stuckert, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2009-09-15

    The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO{sub 2} pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI-Switzerland and AEKI-Hungary. Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g

  20. 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.

  1. Preliminary safety analysis for offgas treatment system of DUPIC fuel manufacturing process at DFDF

    Energy Technology Data Exchange (ETDEWEB)

    Shin, J. M.; Lee, H. H.; Park, J. J.; Yang, M. S

    2000-09-01

    DUPIC fuel fabrication process is a dry processing technology to manufacture CANDU compatible fuel through a direct refabrication process from spent PWR fuel. DUPIC fuel fabrication process consists of the slitting of the spent PWR fuel rods, OREOX processing, homogeneous mixing, pelletizing and sintering. All these processes should be conducted by remote means in a M6 hot cell at IMEF. Since there is a lot of highly radioactive spent fuel(200 kg) to be used in DUPIC fuel fabrication process, safety analysis on DFDF facility is very important to improve the safety of hot cell and to reduce the dose exposure to operator. This report describes the design of IMEF facility, manufacturing equipment and process, offgas treatment system necessary for DUPIC fuel manufacturing process. Also, it provides the flow chart of arising and activity for each nuclide in offgas treatment system and final arising and activity for gaseous waste discharged from offgas treatment equipment into inside of M6 cell during OREOX and sintering processes in DUPIC fuel manufacturing process.

  2. Calculation of the internal pressure of fuel rod from measurements of krypton-85 at its plenum; Calculo de la presion interna de barra combustible a partir de la medida de kripton-85 en su plenum

    Energy Technology Data Exchange (ETDEWEB)

    Arana, I.; Doncel, N.; Casado, C.

    2012-07-01

    ENUSA carried out numerous campaigns of measurement internal pressure of fuel rod irradiated. All of them have been performed of form destructively in a hot cell laboratory which implies a time high to obtain results and a high economic cost to obtain a single data by rod, representative of the end of the irradiation. The objective of the project is to develop a non-destructive measurement and a methodology for reliable calculation that eliminates these problems.

  3. Rod consolidation at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1986-12-01

    A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab.

  4. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

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

    Science.gov (United States)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  6. Evaluation of the thermal-mechanic performance of fuel rods MOX in fuel assemblies 10 x 10; Evaluacion del desempeno termo-mecanico barras combustibles MOX en ensambles combustible 10 x 10

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H., E-mail: hector.hernandez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In the Instituto Nacional de Investigaciones Nucleares (Mexico) , we have been working in proposals of fuel assemblies that bear to the reduction of the plutonium inventories that exist a global level, plutonium coming from the dismantlement of the nuclear weapons as of the one used as fuel inside the reactors in operation at the present time. For this reason besides carrying out the evaluation of the neutron performance is necessary to realize the evaluation of the thermal-mechanic behavior of the rods that compose a fuel assembly with the purpose of determining if under the operation conditions to those that are subjected the fuel does not surpass the limit established and this causes a failure in the fuel element. In this sense when carrying out the analysis of an fuel element of mixed oxides in an arrangement 10 x 10 is observed that under the established operation conditions for the proposed cycle values that surpass the limit established for fuel failure are not presented, therefore the proposed assembly can be used as reload element in the nuclear power plant of Laguna Verde. (Author)

  7. NEUTRONICS STUDIES OF URANIUM-BASED FULLY CERAMIC MICRO-ENCAPSULATED FUEL FOR PWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL

    2012-01-01

    This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature

  8. Development of equipment for fabricating DUPIC fuel powder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Ho; Yang, M. S.; Park, J. J.; Lee, J. W.; Kim, J. H.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Na, S. H

    1999-06-01

    The powder fabrication processes, as the first stage of manufacturing DUPIC (Direct Use of PWR spent fuel In CANDU) fuel, consist of the slitting of spent PWR fuel rods, REOX (Oxidation and REduction of Oxide Fuels) processing to produce the powder feedstock, the milling of the produced powder, the granulation of the milled powder, and the mixing of the granulated powder with pressing lubricants. All these processes should be conducted by remote means in a hot-cell environment where the direct human access is limited to the strictest minimum due to the high radioactivity. This report describe the development of the equipment for fabricating DUPIC fuel powder. These equipment are Slitting Machine, Oxidation and Reduction (OREOX) Furnace, Mill, Roll Compactor, and Mixer. Remote design concept was applied to all the equipment for use in the M6 hot-cell of the IMEF. Mechanical design considerations and capabilities of the equipment for remote operation and maintenance are presented. First prototypes were developed and installed in the DUPIC full scale mock-up and tested using a master-slave manipulator. Redesign and reconstruction were made on each equipment based on mock-up test results. The remote technology acquired through this research was utilized in developing other equipment for DUPIC fuel fabrication, thereby improving safety and increasing productivity. This technology could also be extended to the area of remote handling equipment development for use in hazardous environments. (author). 14 refs., 9 tabs., 21 figs.

  9. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  10. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  11. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Energy Technology Data Exchange (ETDEWEB)

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.

    1999-05-15

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  12. ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

    Directory of Open Access Journals (Sweden)

    Endiah Puji Hastuti

    2015-04-01

    yang sangat lama dan membutuhkan memori yang besar. Kata kunci: aliran turbulen, kanal PWR, CFD, tunak, transien   Coolant flow turbulence on heat transfer process serves to enhance the heat transfer coefficient, likewise flow in the fuel sub channel. Computational fluid dynamic program, FLUENT is a computational program based on finite element, that is able to predict and analyze the dynamics of fluid flow phenomena, accurately. CFD calculation program is selected in this study because of its accurately and it also can provide good visualization. Purpose of this research was to understand the characteristics of heat transfer, mass and momentum of the fuel rod to the coolant visually on: the temperature field, pressure field, and the kinetic energy field, as a function of the flow dynamics within fuel channel, on steady state and transient condition. Analysis of flow dynamics in the fuel channel base on CFD was done by using the PWR sample data with reactor power of 1000 MWe on 17x17 array of fuel. To examine the sensitivity of the flow equation in accordance with the model of turbulent flow on fuel channel, the turbulence equation model of k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, and Reynold stress model (RSM for steady state was used, while for transient turbulence model DES and LES are applied. In the sensitivity analysis of turbulent flow, hexahedral mesh model of three cell geometry each are 0.5 mm, 0.2 mm and 0.15 mm, was selected. The analysis shows that there are similar results of turbulen model Ƙ-ε and Ƙ-ω standard, on steady state analysis. Comparing with Dittus Boelter criteria for Nusselt number, the Reynolds stress model (RSM is recommended. Sensitivity analysis of mesh geometry between cell size 0.5 mm, 0.2 mm and 0.15 mm, indicating that the cell size of 0.5 mm was sufficient. Developed flow already reached on DES and LES model, however only for short time (3 seconds for transient condition. LES model need very long computation time and big memory

  13. A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible.

  14. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Elbel, H.

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

  15. Investigating hydrodynamic characteristics and peculiarities of the coolant flow behind a spacer grid of a fuel rod assembly of the floating nuclear power unit

    Science.gov (United States)

    Dmitriev, S. M.; Doronkov, D. V.; Legchanov, M. A.; Pronin, A. N.; Solncev, D. N.; Sorokin, V. D.; Hrobostov, A. E.

    2016-05-01

    The results of experimental investigations of local hydrodynamics of a coolant flow in fuel rod assembly (FA) of KLT-40C reactor behind a plate spacer grid have been presented. The investigations were carried out on an aerodynamic rig using the gas-phase diffusive tracer test. An analysis of spatial distribution of absolute flow velocity projections and distribution of tracer concentration allowed specifying a coolant flow pattern behind the plate spacer grid of the FA. On the basis of obtained experimental data the recommendations were provided to specify procedures for determining the coolant flow rates for the programs of cell-wise calculation of a core zone of KLT-40C reactor. Investigation results were accepted for the practical use in JSC "OKBM Afrikantov" to assess heat engineering reliability of cores of KLT-40C reactor and were included in a database for verification of CFD programs (CFD-codes).

  16. Application of fully ceramic microencapsulated fuels in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, C.; George, N.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States); Godfrey, A.; Terrani, K.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01

    This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

  17. Preparation of carbon alloy catalysts for polymer electrolyte fuel cells from nitrogen-containing rigid-rod polymers

    Energy Technology Data Exchange (ETDEWEB)

    Chokai, Masayuki [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Integrative Technology Research Institute, Teijin Ltd., 4-3-2, Asahigaoka, Hino, Tokyo 191-8512 (Japan); Taniguchi, Masataka; Shinoda, Tsuyoshi; Nabae, Yuta; Kuroki, Shigeki; Hayakawa, Teruaki; Kakimoto, Masa-aki [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Moriya, Shogo; Matsubayashi, Katsuyuki [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Business Development Division, Nisshinbo Holdings, Inc., 1-2-3, Onodai, Midori-ku, Chiba 267-0056 (Japan); Ozaki, Jun-ichi [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Department of Nanomaterial Systems, Graduate School of Engineering, Gunma University, 1-5-1, Tenjin-cho, Kiryu, Gunma 376-8515 (Japan); Miyata, Seizo [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); New Energy and Industrial Technology Development Organization, 1310 Omiya-cho, Saiwai-ku, Kawasaki, Kanagawa 212-8554 (Japan)

    2010-09-15

    Carbon alloy catalysts (CAC), non-precious metal catalysts for the oxygen reduction reaction (ORR), were prepared from various kinds of nitrogen-containing rigid-rod aromatic polymers, polyimides, polyamides and azoles, by carbonization at 900 C under nitrogen flow. The catalytic activity for ORR was evaluated by the onset potential, which was taken at a current density of -2 {mu}A cm{sup -2}. Carbonized polymers having high nitrogen content showed higher onset potential. In particular, CACs derived from azole (Az5) had an onset potential of 0.8 V, despite being was prepared without any metals. (author)

  18. Partially-reflected water-moderated square-piteched U(6.90)O2 fuel rod lattices with 0.67 fuel to water volume ratio (0.800 CM Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Harms, Gary A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O2 fuel rods.

  19. PWR Cross Section Libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, Carolyn [Texas A& M University; Ilas, Germina [ORNL

    2012-01-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  20. Operational modal analysis of flow-induced vibration of nuclear fuel rods in a turbulent axial flow

    Energy Technology Data Exchange (ETDEWEB)

    De Pauw, B., E-mail: bdepauw@vub.ac.be [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium); Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Weijtjens, W.; Vanlanduit, S. [Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Van Tichelen, K. [Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Berghmans, F. [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium)

    2015-04-01

    Highlights: • We describe an analysis technique to evaluate nuclear fuel pins. • We test a single fuel pin mockup subjected to turbulent axial flow. • Our analysis is based on operational modal analysis (OMA). • The accuracy and precision of our method is higher compared to traditional methods. • We demonstrate the possible onset of a fluid-elastic instability. - Abstract: Flow-induced vibration of nuclear reactor fuel pins can result in mechanical noise and lead to failure of the reactor's fuel assembly. This problem can be exacerbated in the new generation of liquid heavy metal fast reactors that use a much denser and more viscous coolant in the reactor core. An investigation of the flow-induced vibration in these particular conditions is therefore essential. In this paper, we describe an analysis technique to evaluate flow-induced vibration of nuclear reactor fuel pins subjected to a turbulent axial flow of heavy metal. We deal with a single fuel pin mockup designed for the lead–bismuth eutectic (LBE) cooled MYRRHA reactor which is subjected to similar flow conditions as in the reactor core. Our analysis is based on operational modal analysis (OMA) techniques. We show that the accuracy and precision of our OMA technique is higher compared to traditional methods and that it allows evaluating the evolution of modal parameters in operational conditions. We also demonstrate the possible onset of a fluid-elastic instability by tracking the modal parameters with increasing flow velocity.

  1. Direct Measurement of U235 and Pu239 in Spent Fuel Rods with Gamma-Ray Mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Ziock, Klaus-Peter [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Alameda, J. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brejnholt, N. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Decker, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Descalle, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fernandez-Perea, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hill, R. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kisner, R. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Melin, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, B. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ruz, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Soufli, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-09-30

    The amounts of fissile Pu and U in spent nuclear fuel are of primary concern to the safeguards community. In particular, there are issues when safeguards transitions from an item accountancy basis (such as fuel bundles) to a fissile material mass basis as occurs when spent fuel enters a reprocessing plant. Discrepancies occur because item accountancy requires estimating the content of fissile material using indirect techniques such as the fuel burn-up and item-level measurements of radiation emissions from fission by-products. Direct measurement of the fissile content by monitoring line emissions from fissile species themselves is impossible because the lines are much weaker than those emitted by shorter-lived isotopes in the fuel. The goal of this project is to develop a technique to directly measure these weaker lines despite the presence of overwhelming radiation from other isotopes. This is achieved by using gamma-ray mirrors as a narrow band-pass filter. The mirrors reflect only energies of interest toward a HPGe detector that is shielded from direct view of the spent fuel and its fierce emissions. This can significantly improve the reliability with which the mass of fissile material is tracked.

  2. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  3. Assessment of PWR plutonium burners for nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, A J; Shapiro, N L

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  4. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code; Evaluacion del desempeno termomecanico de barras de combustible de un reactor BWR durante una rampa de potencia utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R.

    2010-07-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  5. OECD/NEA Sandia Fuel Project phase I: Benchmark of the ignition testing

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina, E-mail: martina_adorni@hotmail.it [UNIPI (Italy); Herranz, Luis E. [CIEMAT (Spain); Hollands, Thorsten [GRS (Germany); Ahn, Kwang-II [KAERI (Korea, Republic of); Bals, Christine [GRS (Germany); D' Auria, Francesco [UNIPI (Italy); Horvath, Gabor L. [NUBIKI (Hungary); Jaeckel, Bernd S. [PSI (Switzerland); Kim, Han-Chul; Lee, Jung-Jae [KINS (Korea, Republic of); Ogino, Masao [JNES (Japan); Techy, Zsolt [NUBIKI (Hungary); Velazquez-Lozad, Alexander; Zigh, Abdelghani [USNRC (United States); Rehacek, Radomir [OECD/NEA (France)

    2016-10-15

    Highlights: • A unique PWR spent fuel pool experimental project is analytically investigated. • Predictability of fuel clad ignition in case of a complete loss of coolant in SFPs is assessed. • Computer codes reasonably estimate peak cladding temperature and time of ignition. - Abstract: The OECD/NEA Sandia Fuel Project provided unique thermal-hydraulic experimental data associated with Spent Fuel Pool (SFP) complete drain down. The study conducted at Sandia National Laboratories (SNL) was successfully completed (July 2009 to February 2013). The accident conditions of interest for the SFP were simulated in a full scale prototypic fashion (electrically heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate severe accident code validation and to reduce modeling uncertainties within the codes. Phase I focused on axial heating and burn propagation in a single PWR 17 × 17 assembly (i.e. “hot neighbors” configuration). Phase II addressed axial and radial heating and zirconium fire propagation including effects of fuel rod ballooning in a 1 × 4 assembly configuration (i.e. single, hot center assembly and four, “cooler neighbors”). This paper summarizes the comparative analysis regarding the final destructive ignition test of the phase I of the project. The objective of the benchmark is to evaluate and compare the predictive capabilities of computer codes concerning the ignition testing of PWR fuel assemblies. Nine institutions from eight different countries were involved in the benchmark calculations. The time to ignition and the maximum temperature are adequately captured by the calculations. It is believed that the benchmark constitutes an enlargement of the validation range for the codes to the conditions tested, thus enhancing the code applicability to other fuel assembly designs and configurations. The comparison of

  6. Modelling the cracking of pressurised water reactor fuel pellets and its consequences on the mechanical behaviour of the fuel rod; Etude de l'impact de la fissuration des combustibles nucleaires oxyde sur le comportement normal et incidentel des crayons combustible

    Energy Technology Data Exchange (ETDEWEB)

    Helfer, Th

    2006-03-15

    This thesis aims to model the cracking of pressurised water reactor fuel pellets and its consequences on the mechanical behaviour of the fuel rod. Fuel cracking has two main consequences. It relieves the stress in the pellet, upon which the majority of the mechanical and physico-chemical phenomena are dependent. It also leads to pellet fragmentation. Taking fuel cracking into account is therefore necessary to adequately predict the mechanical loading of the cladding during the course of an irradiation. The local approach to fracture was chosen to describe fuel pellet cracking. Practical considerations brought us to favour a quasi-static description of fuel cracking by means of a local damage models. These models describe the appearance of cracks by a local loss of rigidity of the material. Such a description leads to numerical difficulties, such as mesh dependency of the results and abrupt changes in the equilibrium state of the mechanical structure during unstable crack propagations. A particular attention was paid to these difficulties because they condition the use of such models in engineering studies. This work was performed within the framework of the ALCYONE fuel performance package developed at CEA/DEC/SESC which relies on the PLEIADES software platform. ALCYONE provides users with various approaches for modelling nuclear fuel behaviour, which differ in terms of the type geometry considered for the fuel rod. A specific model was developed and implemented to describe fuel cracking for each of these approaches. The 2D axisymmetric fuel rod model is the most innovative and was particularly studied. We show that it is able to assess, thanks to an appropriate description of fuel cracking, the main geometrical changes of the fuel rod occurring under normal and off-normal operating conditions. (author)

  7. Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, C.E.; Cunningham, M.E.; Lanning, D.D. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessed against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.

  8. Monopolar fuel cell stack coupled together without use of top or bottom cover plates or tie rods

    Science.gov (United States)

    Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor)

    2009-01-01

    A monopolar fuel cell stack comprises a plurality of sealed unit cells coupled together. Each unit cell comprises two outer cathodes adjacent to corresponding membrane electrode assemblies and a center anode plate. An inlet and outlet manifold are coupled to the anode plate and communicate with a channel therein. Fuel flows from the inlet manifold through the channel in contact with the anode plate and flows out through the outlet manifold. The inlet and outlet manifolds are arranged to couple to the inlet and outlet manifolds respectively of an adjacent one of the plurality of unit cells to permit fuel flow in common into all of the inlet manifolds of the plurality of the unit cells when coupled together in a stack and out of all of the outlet manifolds of the plurality of unit cells when coupled together in a stack.

  9. Simulation of Thermopower Influence on Fuel Core of Power Rod in Nuclear Power Plant (NPP Active Zone

    Directory of Open Access Journals (Sweden)

    I. S. Kulikov

    2010-01-01

    Full Text Available The paper considers problems of modern methods for  calculation of designs and materials of nuclear power. A model of numerical analysis for stress-strain state of fuel pins in the NPP active zone is proposed in the paper. The paper contains simulation concerning a fuel core section of a nuclear reactor heat-generating element with subsequent solution of a temperature and thermoelastic problem in computer program complex FEA ANSYS Workbench 11.0. All the obtained results have passed through checking procedure.

  10. Direct measurement of 235U in spent fuel rods with Gamma-ray mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Ruz, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brejnholt, N. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Alameda, J. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Decker, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Descalle, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fernandez-Perea, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hill, R. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kisner, R. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Melin, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, B. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soufli, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ziock, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pivovaroff, M. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-22

    We report here that direct measurement of plutonium and uranium X-rays and gamma-rays is a highly desirable nondestructive analysis method for the use in reprocessing fuel environments. The high background and intense radiation from spent fuel make direct measurements difficult to implement since the relatively low activity of uranium and plutonium is masked by the high activity from fission products. To overcome this problem, we make use of a grazing incidence optic to selectively reflect Kα and Kβ fluorescence of Special Nuclear Materials (SNM) into a high-purity position-sensitive germanium detector and obtain their relative ratios.

  11. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1980-November 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-02-01

    Four tasks are reported: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  12. Irradiation and lithium presence influence on the crystallographic nature of zirconia in the framework of PWR zircaloy 4 fuel cladding corrosion study; Influence de l'irradiation et de la presence du lithium sur la nature cristallographique de la zircone dans le cadre de l'etude de la corrosion du zircaloy 4 en milieu reacteur a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Gibert, C

    1999-07-01

    The-increasing deterioration of the initially protective zirconia layer is one of the hypotheses which can explain the impairment with time of PWR fuel cladding corrosion. This deterioration could be worsened by irradiation or lithium presence in the oxidizing medium. The aim of this thesis was to underline the influence of those two parameters on zirconia crystallographic nature. We first studied the impact of ionic irradiation on pure, powdery, monoclinic zirconia and oxidation formed zirconia, mainly with X-ray diffraction and Raman microscopy. The high or low energy particles used (Kr{sup n+-}, Ar{sup n+}) respectively favored electronic or atomic defaults production. The crystallographic analyses showed that these irradiation have a significant effect on zirconia by inducing nucleation or growth of tetragonal phase. The extent depends on sample nature and particles energy. In all cases, phase transformation is correlated with crystalline parameters, grain size and especially micro-stress changes. The results are consistent with those obtained with 1 to 5 cycles PWR claddings. Therefore, the corrosion acceleration observed in reactor can partly be explained by the stress fields appearance under irradiation, which is particularly detrimental to zirconia layer cohesion. Last, we have underlined that the presence of considerable amounts of lithium in the oxidizing medium ((> 700 ppm) induces the disappearance of the tetragonal zirconia located at the metal/oxide interface and the appearance of a porosity of the dense under layer, which looses its protectiveness. (author)

  13. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  14. Analysis of the behavior of irradiated BWR fuel rod in storage dry conditions; Analisis del comportamiento de una barra combustible irradiada BWR en condiciones de almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Montes, D.; Ruiz-Hervias, J.; Munoz-Reja, C.

    2014-07-01

    In order to complete previous studies of creep on PWR sheath material, developed a joint experimental program by CSN, ENRESA and ENUSA about BWR (Zircaloy-2) sheath material. This program consisted in creep tests and then on the material under creep, compression testing diametral obtaining the permissible displacement of the sheath to break. (Author)

  15. Morphological analysis of zirconium nuclear fuel retaining rods braided with SiC: Quality assurance and defect identification

    Science.gov (United States)

    Glazoff, Michael V.; Hiromoto, Robert; Tokuhiro, Akira

    2014-08-01

    In the after-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Among the methods explored currently to improve zircaloys’ thermal stability in off-normal conditions, using a protective coat of the SiC filaments is considered because silicon carbide is well known for its remarkable chemical inertness at high temperatures. A typical SiC fiber contains ∼50,000 individual filaments of 5-10 μm in diameter. In this paper, an effort was made to develop and apply mathematical morphology to the process of automatic defect identification in Zircaloy-4 rods braided with the protective layer of the silicon carbide filament. However, the issues of the braiding quality have to be addressed to ensure its full protective potential. We present the original mathematical morphology algorithms that allow solving this problem of quality assurance successfully. In nuclear industry, such algorithms are used for the first time, and could be easily generalized to the case of automated continuous monitoring for defect identification in the future.

  16. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  17. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Science.gov (United States)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  18. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  19. Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities.

    Science.gov (United States)

    Ruz, J; Descalle, M A; Alameda, J B; Brejnholt, N F; Chichester, D L; Decker, T A; Fernandez-Perea, M; Hill, R M; Kisner, R A; Melin, A M; Patton, B W; Soufli, R; Trellue, H; Watson, S M; Ziock, K P; Pivovaroff, M J

    2016-06-01

    The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. The experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in future measurement campaigns.

  20. Study of power peak migration due to insertion of control bars in a PWR reactor; Estudo da migracao do pico de potencia em funcao da insercao das barras de controle em um reator refrigerado a agua

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Costa, Danilo Leite; Borges, Diogo da Silva; Lava, Deise Diana; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: danilolc26@gmail.com, E-mail: diogosb@outlook.com, E-mail: deisedy@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown.

  1. 棒束燃料组件特征栅元CFD方法研究%CFD Method Research on Characteristic Cells in Rod Bundle Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    陈杰; 陈炳德; 张虹

    2011-01-01

    Two characteristic cells are in AFA-3G fuel assembly, that is typical cell and control rod guide cell. And there are some rules on the arrangement of mixing vanes. For the two characteristic cells, mixing capability is evaluated axially from the point of the first and second kind of sub-channel with CFD method.Mass mixing and heat mixing are interaction but different with each other. Although the mass mixing in the first kind of sub-channel is stronger, the thermal capability of the two is to some tune from the point of heat transfer. In the experiment research on thermal-hydraulic performance of AFA-3G fuel assembly, the arrangements of mixing vanes should refer to the two spacer grids of characteristic cells.%AFA-3G燃料组件中存在典型栅元和控制棒导向管栅元两种特征栅元,定位格架搅混翼的排列也具有一定的规律性.本文采用计算流体力学(CFD)方法,分别针对两种特征栅元,从第一类子通道和第二类子通道的角度,沿程评价其交混性能.质量交混与热交混紧密联系又相互区别,第一类子通道质量交换较强,但从传热角度,二者性能相当.AFA-3G燃料组件热工水力性能的实验研究中,格架搅混翼的排列方式应分别参照两种特征栅元格架.

  2. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods (1.506-cm Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed from 1962–1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967.a The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, relative fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector” (see Reference 1). The experiment studied in this evaluation was the second of the series and had the fuel rods in a 1.506-cm-triangular pitch. One critical configuration was found (see Reference 3). Once the critical configuration had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U,bc and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configuration are described in Sections 1.3, 1.4, and 1.7, respectively.

  3. DNBR Analysis of a Dual-Cooled Annular Fuel for the OPR1000 Application

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; In, W. K.; Oh, D. S.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    A dual-cooled annular fuel for a pressurized water reactor (PWR) has been introduced for a significant amount of reactor power uprate. The Korea Atomic Energy Research Institute (KAERI) has been performing a research to develop a dual-cooled annular fuel for the power uprate of 20% in an optimized PWR in Korea, OPR1000. Several thermal-hydraulic tasks exist for the application of the dual-cooled annular fuel to OPR1000. The primary task is the balance of the minimum DNBR (MDNBR) between the inner and outer channels since the coolant flows through the circular inner channel of annular fuel as well as the outer subchannels formed between the fuel rods. The MDNBR balance has been known to largely depend on the thermal conductance in the inner and outer gaps. Another task is to evaluate the operating condition that the inner coolant channel is partially blocked. This study calculated the MDNBR in the inner and outer channels depending on the thermal gap conductance, i.e., inner and outer gap width. The acceptable range of gap width is determined for the MDNBR not to exceed the DNBR limit during anticipated operational occurrences (AOOs) as well as normal operation. The limit for the flow blockage in the inner channel is also estimated based on the DNBR analysis

  4. Neutronic evaluation of thorium and reprocessed fuels by GANEX and UREX+ in ADS

    Energy Technology Data Exchange (ETDEWEB)

    Barros, Graiciany, E-mail: graiciany.barros@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., 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

    A conceptual design of accelerator driven systems (ADS) that utilize thorium and reprocessed fuel in order to produce {sup 233}U and to transmute high radiotoxicity isotopes in spent nuclear fuel has been proposed. The use of thorium and reprocessed fuel in an ADS is one of the clean, safe, and economical solutions for the problem of nuclear waste. In this study, the aim was to compare the neutronic behavior of the core using spent fuel reprocessed by GANEX (Group ActiNide EXtraction) and UREX+ (Uranium Extraction), both spiked with thorium. The simulated design was a cylinder fuelled with a hexagonal lattice with 156 fuel rods. One of the studied fuels was a mixture based upon Pu-MA, removed from PWR-spent fuel, theoretically reprocessed by GANEX reprocessing and spiked with 82% of thorium. The other fuel was a reprocessed fuel obtained theoretically from UREX+ (Uranium Extraction) process and spiked with 82% of thorium. Monteburns 2.0 (MCNP5/ORIGEN 2.1) code was used to simulate the neutronic aspects of the fuels. The multiplication factors, the neutron spectra, and the nuclear fuel evolution were analyzed during 10 years of burn-up. The results allowed comparing the two reprocessing techniques, the {sup 233}U production and the reduction in the amount of high radiotoxicity isotopes of these fuels. (author)

  5. Experimental data report for test TS-2; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1993-01-01

    本報告書は、1990年2月に実施した照射済BWR燃料を用いた2回目の反応度事故模擬実験であるTS-2について実験データをまとめたものである。TS-2実験に使用した試験燃料は初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3Gwd/tであった。NSRRにおける照射実験は、大気圧、室温の静止水冷却条件下で行い、発熱量は72pm5cal/g・fuel(ピークエンタルピ66pm5cal/g・fuel)を与えた。その結果燃料破損は生じなかった。実験条件、実験方法、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  6. Experimental data report for test TS-1; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1992-01-01

    本報告書は、1989年10月に実施した照射済BWR燃料を用いた最初の反応度事故模擬実験であるTS-1について、実験データをまとめたものである。TS-1実験に使用した試験燃料は、初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3GWd/tであった。NSRRにおける照射実験は、新たに開発した専用の2重カプセルを用い、大気圧・室温の静止水冷却条件下で行い、発熱量61cal/g・fuel(ピークエンタルピ55cal/g・fuel)を与えた。その結果、燃料破損は生じなかった。実験条件、実験方法、燃料燃焼度の測定結果、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  7. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  8. SCOR 1000: an economic and innovative conceptual design PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)

    2007-07-01

    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  9. Morphoelastic rods

    CERN Document Server

    Tiero, Alessandro

    2014-01-01

    We propose a mechanical theory describing elastic rods which, like plant organs, can grow and can change their intrinsic curvature and torsion. The equations ruling accretion and remodeling are obtained by combining balance laws involving non-standard forces with constitutive prescriptions filtered by a dissipation principle that takes into account both standard and non-standard working.

  10. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type; Calculo de las razones de generacion de calor lineal que violen el limite termomecanico de deformacion plastica de la camisa en funcion del quemado de una barra combustible tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2003-07-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  11. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yang, M. S.; Park, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    This study describes the DUPIC fuel development program in KAERI as follows; Burning spent PWR fuel again in CANDU by DUPIC, Compatibility with existing CANDU system, Feasibility of DUPIC fuel fabrication, Waste reduction, Safeguard ability, Economics of DUPIC fuel cycle, The DUPIC fuel development program, and International prospective. 5 refs., 10 figs.

  12. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement

    1999-04-01

    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  13. CONTROL ROD

    Science.gov (United States)

    Zinn, W.H.; Ross, H.V.

    1958-11-18

    A control rod is described for a nuclear reactor. In certaln reactor designs it becomes desirable to use a control rod having great width but relatively llttle thickness. This patent is addressed to such a need. The neutron absorbing material is inserted in a triangular tube, leaving volds between the circular insert and the corners of the triangular tube. The material is positioned within the tube by the use of dummy spacers to achleve the desired absorption pattern, then the ends of the tubes are sealed with suitable plugs. The tubes may be welded or soldered together to form two flat surfaces of any desired width, and covered with sheetmetal to protect the tubes from damage. This design provides a control member that will not distort under the action of outside forces or be ruptured by gases generated within the jacketed control member.

  14. Study of a criticality accident involving fuel rods and water outside a power reactor; Etude d'un accident de criticite mettant en presence des crayons combustibles et de l'eau hors reacteur de puissance

    Energy Technology Data Exchange (ETDEWEB)

    Beloeil, L

    2000-05-30

    It is possible to imagine highly unlikely but numerous accidental situations where fuel rods come into contact with water under conditions close to atmospheric values. This work is devoted to modelling and simulation of first instants of the power excursion that may result from such configurations. We show that void effect is a preponderant feedback for most severe accidents. The formation of a vapour film around the rods is put forward and confirmed with the help of experimental transients using electrical heating. We propose then a vapour/liquid flow model able to reproduce void fraction evolution. The vapour film is treated as a compressible medium. Conservation balance equations are solved on a moving mesh with a two-dimensional scheme and boundary conditions taking notice of interfacial phenomena and axial escape possibility. Movements of the liquid phase are modelled through a non-stationary integral equation and a dissipative term suited to the particular geometry of this flow. The penetration of energy into the liquid is also calculated. Thus, the coupling of aerodynamic and hydrodynamic modules gives results in excellent agreement with experiments. Next, neutronic phenomena into the fuel pellet, their feedback effects and the distribution of power through the rod are numerically translated. For each developed module, validation tests are provided. Then, it is possible to simulate the first seconds of the whole criticality accident. Even if this calculation tool is only a way of study as a first approach, performed simulations are proving coherent with reported data on recorded accidents. (author)

  15. Analysis of Subchannel and Rod Bundle PSBT Experiments with CATHARE 3

    Directory of Open Access Journals (Sweden)

    M. Valette

    2012-01-01

    Full Text Available This paper presents the assessment of CATHARE 3 against PWR subchannel and rod bundle tests of the PSBT benchmark. Noticeable measurements were the following: void fraction in single subchannel and rod bundle, multiple liquid temperatures at subchannel exit in rod bundle, and DNB power and location in rod bundle. All these results were obtained both in steady and transient conditions. Void fraction values are satisfactory predicted by CATHARE 3 in single subchannels with the pipe module. More dispersed predictions of void values are obtained in rod bundles with the CATHARE 3 3D module at subchannel scale. Single-phase liquid mixing tests and DNB tests in rod bundle are also analyzed. After calibrating the mixing in liquid single phase with specific tests, DNB tests using void mixing give mitigated results, perhaps linked to inappropriate use of CHF lookup tables in such rod bundles with many spacers.

  16. Neutron Flux Depression in the UO{sub 2}-PuO{sub 2}(15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment; Depresion de flujo neutronico en las barras combustibles de UO2-PuO2(15 al 30%) del experimento de irradiacion IVO-FR2-Vg7

    Energy Technology Data Exchange (ETDEWEB)

    Lopez, J.; Fernandez, J. L.

    1983-07-01

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO{sub 2}-PUO{sub 2} (15 to 30% PUO{sub 2}) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs.

  17. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees o