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Sample records for burnup nuclear fuel

  1. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

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

  2. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

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

  3. Fuel burnup monitor for nuclear reactors

    International Nuclear Information System (INIS)

    An in-service detector is designed using the principle of comparing temperatures in the fuel element and in the detector material. The detector consists of 3 metallic heat conductors insulated with ceramic insulators, two of them with uranium fuel spheres at the end. One sphere is coated with zirconium, the other with zirconium and gold. The precision of measurement of the degree of fuel burnup depends on the precision of the measurement of temperature and is determined from the difference in temperature gradients of the two uranium fuel spheres in the detector. (M.D.)

  4. Automated system for determining the burnup of spent nuclear fuel

    Directory of Open Access Journals (Sweden)

    Mokritskii V. A.

    2014-12-01

    Full Text Available The authors analyze their experience in application of semi-conductor detectors and development of a breadboard model of the monitoring system for spent nuclear fuel (SNF. Such system should use CdZnTe-detectors in which one-charging gathering conditions are realized. The proposed technique of real time SNF control during reloading technological operations is based on the obtained research results. Methods for determining the burnup of spent nuclear fuel based on measuring the characteristics of intrinsic radiation are covered in many papers, but those metods do not usually take into account that the nuclear fuel used during the operation has varying degrees of initial enrichment, or a new kind of fuel may be used. Besides, the known methods often do not fit well into the existing technology of fuel loading operations and are not suitable for operational control. Nuclear fuel monitoring (including burnup determination system in this research is based on the measurement of the spectrum of natural gamma-radiation of irradiated fuel assemblies (IFA, as from the point of view of minimizing the time spent, the measurement of IFA gamma spectra directly during fuel loading is optimal. It is the overload time that is regulated rather strictly, and burnup control operations should be coordinated with the schedule of the fuel loading. Therefore, the real time working capacity of the system should be chosen as the basic criterion when constructing the structure of such burnup control systems.

  5. Nuclear fuel burn-up credit for criticality safety justification of spent nuclear fuel storage systems

    International Nuclear Information System (INIS)

    Burn-up credit analysis of RBMK-1000 an WWER-1000 spent nuclear fuel accounting only for actinides is carried out and a method is proposed for actinide burn-up credit. Two burn-up credit approaches are analyzed, which consider a system without and with the distribution of isotopes along the height of the fuel assembly. Calculations are performed using SCALE and MCNP computer codes

  6. Instrumentation for measuring the burnup of spent nuclear fuel

    International Nuclear Information System (INIS)

    Many different methods or procedures have been developed to measure reactivity of fissil materials. Few of these, however, have been designed specifically for light water reactor fuel or have actually been used to measure the reactivity of spent fuel. The methods that have been used to make measurements of related systems are the 252Cf source-driven noise analysis method, a noise analysis method using natural neutron sources, subcritical assembly measurements, and pulsed neutron techniques. Several different approaches to directly measuring burnup have been developed by various organizations. The experimental work on actual spent nuclear fuel utilizing reactivity measurement techniques is insufficient to provide conclusive evidence of the applicability of these techniques for verifying fuel burnup. The work with burnup meters indicates, however, that good correlations can be obtained with any of the systems. A burnup meter's primary function would be a secondary assurance that the administrative records are not grossly in error. Reactivity measurements provide information relating to the reactivity of the fuel only under the conditions measured. Criticality prevention design requirements will necessitate that casks accommodate a minimum burnup level for a given initial enrichment (i.e., a maximum reactivity). Direct measurement of the burnup will enable an easy determination of whether a particular fuel assembly can be shipped in a specific cask with a minimum number of additional correlations

  7. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    International Nuclear Information System (INIS)

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  8. Burnup performances of boron nitride and boron coated nuclear fuels

    International Nuclear Information System (INIS)

    The nuclear fuels of urania (UOV) and 5% and 10% gadolinia (Gd2O3) containing UO2 previously produced by sol-gel technique were coated with first boron nitride (BN) then boron (B) thin layer by chemical vapor deposition (CVD) and also by plasma enhanced chemical vapor deposition (PECVD) techniques to increase the fuel cycle length and to improve the physical properties. From the cross-sectional view of BN and B layers taken from scanning electron microscope (SEM), the excellent adherence of BN onto fuel and B onto BN layer was observed in both cases. The behavior of fuel burnup, depletion of BN and B, the effect of coating thickness and also Gd2O3 content on the burnup performances of the fuels were identified by using the code WIMS-D/4 for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) cores. The optimum thickness ratio of B to BN was found as 4 and their thicknesses were chosen as 40 mm and 10 mm respectively in both reactor types to get extended cycle length. The assemblies consisting of fuels with 5% Gd2O3 and also coated with 10 mm BN and 40 mm B layers were determined as candidates for getting higher burnup in both types of reactors

  9. Modelling of fission gas behaviour in high burnup nuclear fuel

    International Nuclear Information System (INIS)

    The safe and economic operation of nuclear power plants (NPPs) requires that the behaviour and performance of the fuel can be calculated reliably over its expected lifetime. This requires highly developed codes that treat the nuclear fuel in a general manner and which take into account the large number of influences on fuel behaviour, in particular the trend of NPP operators to increase the fuel burnup. With higher burnup, more fission events impact the material characteristics of the fuel and significant restructuring can be observed. At local burnups in excess of 60-75 MWd/kgU, the microstructure of nuclear fuel pellets differs markedly from the as-fabricated structure. This high burnup structure (HBS) is characterised by three principal features: 1) low matrix xenon concentration, 2) sub-micron grains and 3) a high volume fraction of micrometer-sized pores. The peculiar features of the HBS affect the fuel performance and safety; the large retention of fission gas within the HBS could lead to significant gas release at high burnups, either through the degradation of thermal conductivity or through direct release. The present work has focussed on the development and evaluation of HBS fission gas transport models, especially on two features: the equilibrium xenon concentration in the matrix of the HBS in UO2 fuel pellets, and the growth of the HBS porosity and its effect on fission gas release. A steady-state fission gas model has been developed to examine the importance of grain boundary diffusion for the gas dynamics in the HBS. It was possible to simulate the ∼0.2 wt% experimentally observed xenon concentration. The value of the grain boundary diffusion coefficient is not important for diffusion coefficient ratios in excess of ∼10”4. The model exhibits a high sensitivity to principally three parameters: the grain diffusion coefficient, the bubble number density and the re-solution rate coefficient. The model can reproduce the observed HBS xenon depletion

  10. An overview of burnup credit application in spent nuclear fuel management

    International Nuclear Information System (INIS)

    The current status of burnup credit application has been overviewed for spent nuclear fuel management. It was revealed that the use of burnup credit is practically limited to spent nuclear fuel storage, for which selected actinides-only are taken into account

  11. Specific behaviour aspects at extended burnup operation of PHWR nuclear fuels

    International Nuclear Information System (INIS)

    In order to evaluate the influence of burnup extension on PHWR nuclear fuel performance, the paper presents and discusses the specific potentially life-limiting factors at extended burnup for this type of fuel using recent experimental evidence and making a direct comparison with LWR fuel performance. (Author)

  12. Impact of extended burnup on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The Advisory Group Meeting was held in Vienna from 2 to 5 December 1991, to review, analyse, and discuss the effects of burnup extension in both light and heavy water reactors on all aspects of the fuel cycle. Twenty experts from thirteen countries participated in this meeting. There was consensus that both economic and environmental benefits are driving forces toward the achievement of higher burnups and that the present trend of burnup extension may be expected to continue. The extended burnup has been considered for the three main stages of the fuel cycle: the front end, in-reactor issues and the back end. Thirteen papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

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

  14. Radiochemical analysis of nuclear fuel burn-up and spent fuel key nuclides

    International Nuclear Information System (INIS)

    Destructive radiochemical analysis of spent nuclear fuels is an important tool to determine burn-up with high accuracy and to better understand the process of depletion and formation of actinides and fission products during irradiation as a result of fission and successive neutron capture. The resulting isotope inventories and nuclear databases that are created, are of high importance to evaluate the performance of nuclear fuels in a reactor, to evaluate computer codes applied for a safe transport, storage and disposal/reprocessing of spent fuels and to safeguard fissile material. The objective is to provide chemical and radiochemical analyses procedures for an accurate determination of isotopic compositions and concentrations of actinides and fission products in different types of industrial (UO2, MOX) and experimental nuclear fuels (UAlx, U3Si2, UMo, ...). For a burn-up determination program typically 21 actinides and fission products are analyzed. For an extended characterization program this can increase to up to approximately 50 isotopes

  15. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    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); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  16. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO2 fuel in Japan. The design concept should be compatible with UO2 fuel design. High burnup UO2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO2 8 x 8 array fuel developed for a second step of UO2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  17. Dry Storage Demonstration for High-Burnup Spent Nuclear Fuel-Feasibility Study

    International Nuclear Information System (INIS)

    Initially, casks for dry storage of spent fuel were licensed for assembly-average burnup of about 35 GWd/MTU. Over the last two decades, the discharge burnup of fuel has increased steadily and now exceeds 45 GWd/MTU. With spent fuel burnups approaching the licensing limits (peak rod burnup of 62 GWd/MTU for pressurized water reactor fuel) and some lead test assemblies being burned beyond this limit, a need for a confirmatory dry storage demonstration program was first identified after the publication in May 1999 of the U.S. Nuclear Regulatory Commissions (NRC) Interim Staff Guidance 11 (ISG-11). With the publication in July 2002 of the second revision of ISG-11, the desirability for such a program further increased to obtain confirmatory data about the potential changes in cladding mechanical properties induced by dry storage, which would have implications to the transportation, handling, and disposal of high-burnup spent fuel. While dry storage licenses have kept pace with reactor discharge burnups, transportation licenses have not and are considered on a case by case basis. Therefore, this feasibility study was performed to examine the options available for conducting a confirmatory experimental program supporting the dry storage, transportation, and disposal of spent nuclear fuel with burnups well in excess of 45 GWd/MTU

  18. Use of axial burnup distribution profile in the nuclear safety analysis of spent nuclear fuel storage for WWER reactors in Ukraine

    International Nuclear Information System (INIS)

    The nuclear safety analysis of spent fuel storages taking into account fuel burnup should allow for burnup distribution along the height of the assembly. We propose a method based on an analysis of the axial burnup profiles of spent fuel assemblies. This method can be used in nuclear safety justification of spent fuel management and storage systems

  19. Non destructive assay of nuclear LEU spent fuels for burnup credit application

    International Nuclear Information System (INIS)

    Criticality safety analysis devoted to spent fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as Burnup credit. To be used, Burnup credit involves obtaining evidence of the reactivity loss with a Burnup measurement. Many non destructive assays (NDA) based on neutron as well as on gamma ray emissions are devoted to spent fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously and the link to Burnup is a power link. As a result, burn-up determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels but the convenience of the gamma spectrometric methods is very dependent on characteristics of spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several Nuclear Power Plants in western Europe, gives the average Burnup within a 5% uncertainty and also the extremity Burnup, 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (Active Neutron Interrogation, Passive Neutron

  20. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P. [Russian Federal Nuclear Center, Snezhinsk (Russian Federation)

    1998-07-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  1. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    International Nuclear Information System (INIS)

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  2. Burnup determination and age dating of spent nuclear fuel using noble gas isotopic analysis

    International Nuclear Information System (INIS)

    During the chopping and dissolving phases of reprocessing, gases (such as tritium, krypton, xenon, iodine, carbon dioxide, nitrogen oxide, and steam) are released. These gases are traditionally transferred to a gas-treatment system for treatment, release, and/or recycle. Because of their chemically inert nature, the xenon and krypton noble gases are generally released directly into the loser atmosphere through the facility's stack. These gases (being fission products) contain information about the fuel being reprocessed and may prove a valuable monitor of reprocessing activities. Two properties of the fuel that may prove valuable from a safeguards standpoint are the fuel burnup and the fuel age (or time since discharge from the reactor). Both can be used to aid in confirming declared activities, and the burnup is generally indicative of the usability of the fuel for fabricating nuclear explosives. A study has been ongoing at Los Alamos National Laboratory to develop a methodology to determine spent-fuel parameters from measured xenon and/or krypton isotopic ratios on-stack at reprocessing facilities. This study has resulted in the generation of the NOVA data analysis code, which links to a comprehensive database of reactor physics parameters (calculated using the Monteburns 3.01 code system). NOVA has been satisfactorily tested for burnup determination of weapons-grade fuel from a US production reactor. Less effort has been spent quantifying NOVA's ability to predict burnup and fuel age for power reactor fuel. The authors describe the results predicted by NOVA for xenon and krypton isotopic ratios measured after the dissolution of spent-fuel samples from the Borssele reactor. The Borssele reactor is a 450-MW(electric) pressurized water reactor (PWR) consisting of 15 x 15 KWU assemblies. The spent-fuel samples analyzed were single fuel rods removed from one assembly and dissolved at the La Hague reprocessing facility. The assembly average burnup was estimated at 32

  3. Chemical analytical considerations on the determination of burnup in irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Burnup in an irradiated nuclear fuel may be defined as the energy produced per mass unit, from the time the fuel is introduced into the reactor and until a given moment. It is usually shown in megawatt/day or megawatt/hour generated per ton or kilo of fuel. It is also indicated as the number of fission produced per volume unit (cm3) or per every 100 initial fissionable atoms. The yield of a power plant is directly related to the burnup of its fuel load and knowing the latter contributes to optimizing the economy in reactor operation and the related technologies. The development of nuclear fuels and the operation of reactors require doing with exact and accurate methods allowing to know the burnup. Errors in this measurement have an incidence upon the fuel design, the physical and nuclear calculations, the shielding requirements, the design of vehicles for the transportation of irradiated fuels, the engineering of processing plants, etc. All these factors, in turn, have an incidence upon the cost of nuclear power generation. (Author)

  4. Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages

    International Nuclear Information System (INIS)

    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

  5. 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

  6. Simulation of the behaviour of nuclear fuel under high burnup conditions

    International Nuclear Information System (INIS)

    Highlights: • Increasing the time of nuclear fuel into reactor generates high burnup structure. • We analyze model to simulate high burnup scenarios for UO2 nuclear fuel. • We include these models in the DIONISIO 2.0 code. • Tests of our models are in very good agreement with experimental data. • We extend the range of predictability of our code up to 60 MWd/KgU average. - Abstract: In this paper we summarize all the models included in the latest version of the DIONISIO code related to the high burnup scenario. Due to the extension of nuclear fuels permanence under irradiation, physical and chemical modifications are developed in the fuel material, especially in the external corona of the pellet. The codes devoted to simulation of the rod behaviour under irradiation need to introduce modifications and new models in order to describe those phenomena and be capable to predict the behaviour in all the range of a general pressurized water reactor. A complex group of subroutines has been included in the code in order to predict the radial distribution of power density, burnup, concentration of diverse nuclides and porosity within the pellet. The behaviour of gadolinium as burnable poison also is modelled into the code. The results of some of the simulations performed with DIONISIO are presented to show the good agreement with the data selected for the FUMEX I/II/III exercises, compiled in the NEA data bank

  7. Key issues in nuclear fuel cycle concerning high burn-up strategy

    International Nuclear Information System (INIS)

    In the present high burn-up strategy in Japan, the economic efficiency and reduction of the spent nuclear fuel have been in progress. On the other hand, in the further progress of the strategy, several issues may appear. The amount and activity of nuclides, heat generation, and radiation for a fuel pin in the typical 17x17 PWR assembly were calculated as functions of burn-up and cooling time, using the SWAT code system. Waste loading in glass waste forms from spent UO2 fuel and MOX fuel were discussed, assuming the number of glass canisters of 150 liter per THM is 1.25 at 45 GWd/THM. The number of glass canisters per GWd is almost constant in the range of burn-up up to 70 GWd/THM. The amount of molybdate from Pu-239 fissions linearly increases as a function of burn-up similarly like increase from U-235 fissions. The current vitrification technology may not face serious situation to be required substantial reduction in waste loading relating to molybdate up to 70 GWd/THM. The initial cooling period prior to vitrification, the waste loading and the interim storage period prior to final disposal are major factors which determine the way of storage and final disposal. The higher burn-up above 45 GWd/THM may require pretreatment of HLLW or substantial reduction in waste loading to retain the integrity of the ceramic melter for e.g. five years. Further promotion of high burn-up strategy should be consistent with nuclear fuel cycle including waste management. A potential approach, a conceptual new reprocessing system for thermal reactors is described. (author)

  8. Burnup credit application in criticality analysis of storage casks with spent RBMK-1500 nuclear fuel

    International Nuclear Information System (INIS)

    Nuclear criticality safety analysis of two types of the casks CASTOR RBMK-1500 and CONSTOR RBMK-1500 was performed using the SCALE 4.3 computer code system. These casks are planned for an interim dry storage of spent nuclear fuel at Ignalina nuclear power plant. Effective neutron multiplication factor keff was calculated for different density of the water inside the casks for unfavorable operational cases and for assumed hypothetical accident conditions when fuel in the system is fresh and fuel is depleted (i.e. burnup credit taken into account). Results show that for all cases effective neutron multiplication factor keff is less then allowable value 0.95. (author)

  9. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    International Nuclear Information System (INIS)

    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, keff, of a spent nuclear fuel package. Fifty-seven UO2, UO2/Gd2O3, and UO2/PuO2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on keff (which can be a function of the trending parameters) such that the biased keff, 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 loading criteria and confirm proper assembly selection prior to loading

  10. Actinide-only burnup credit methodology for PWR spent nuclear fuel

    International Nuclear Information System (INIS)

    A conservative methodology is presented that would allow taking credit for burnup in the criticality safety analysis of spent nuclear fuel (SNF) packages. The method is based on the assumption that the isotopic concentration in the SNF and cross sections of each isotope for which credit is taken must be supported by validation experiments. The method allows credit for 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. 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 spent nuclear fuel 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 by use of UO2 and UO2/Puo2 critical experiments. 3. Establish conditions for the SNF (depletion analysis) and package (criticality analysis) which bounds keff. 4. Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). 5. Verify by measurement that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. (author)

  11. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  12. Evaluation of burnup credit for accommodating PWR spent nuclear fuel in high-capacity cask designs

    International Nuclear Information System (INIS)

    This paper presents an evaluation of the amount of burnup credit needed for high-density casks to transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic 32-assembly cask and the current regulatory guidance were used as bases for this evaluation. By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based loading curves, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of SNF assemblies in high-capacity storage and transportation casks. The impact of varying selected calculational assumptions is also investigated, and considerable improvement in effectiveness is shown with the inclusion of the principal fission products (FPs) and minor actinides and the use of a bounding best-estimate approach for isotopic validation. Given sufficient data for validation, the most significant component that would improve accuracy, and subsequently enhance the utilization of burnup credit, is the inclusion of FPs. (author)

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

    International Nuclear Information System (INIS)

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

  14. Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One

    International Nuclear Information System (INIS)

    Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design

  15. PLD-IDMS studies towards direct measurement of burn-up of nuclear fuel

    International Nuclear Information System (INIS)

    A method based on Pulsed laser deposition followed by Isotope dilution mass spectrometric method is evaluated towards the possibility of direct measurement of burn up of nuclear fuel and also to find out spatial distribution of burn-up along the pellet. The wave length dependent results show larger error with 1064 nm, compared to 532 nm laser beam. Much less error is expected with shorter wave length and shorter pulse width laser beam. Further work is being carried out in this direction

  16. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    International Nuclear Information System (INIS)

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced

  17. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Bang, Je Geon; Kim, Dae Ho; Yang, Yong Sik; Song, Keun Woo

    2007-12-15

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced.

  18. Burnup extension and evolution in the fuel management of EDF's nuclear power plants

    International Nuclear Information System (INIS)

    Today the use of enhanced nuclear fuels that can sustain higher burnups has allowed a better optimization of the fuel management in nuclear power plants. The optimization for the near future is based on 3 aims: -) a better competitiveness of nuclear energy, longer campaigns mean a higher availability and less refueling so it has a direct impact on costs, -) a better flexibility to meet energy demand: a modulation of cycle lengths by more or less 2 months is possible by introducing or withdrawing 8 assemblies in the refueling load, this modulation will allow an optimization of the scheduling of the refueling shutdowns with respect to the seasonal energy demand peaks, -) a reduced volume of spent fuels (but with a higher level of radioactivity). (A.C.)

  19. High burnup fuel development program in Japan

    International Nuclear Information System (INIS)

    A step wise burnup extension program has been progressing in Japan to reduce the LWR fuel cycle cost. At present, the maximum assembly burnup limit of BWR 8 Χ 8 type fuel (B. Step II fuel) is 50GWd/t and a limited numbers of 9 Χ 9 type fuel (B. Step III fuel) with 55GWd/t maximum assembly burnup has been licensed by regulatory agencies recently. Though present maximum assembly burnup limit for PWR fuel is 48GWd/t (P. Step I fuel), the licensing work has been progressing for irradiation testing on a limited number of fuel assemblies with extended burnup of up to 55GWd/t (p. Step II fuel) Design of high burnup fuel and fabrication test are carried out by vendors, and subsequent irradiation test of fuel rods is conducted jointly by utilities and vendors to prepare for licensing. It is usual to make an irradiation test for vectarion, using lead use assemblies by government to confirm fuel integrity and reliability and win the public confidence. Nuclear Power Engineering Corporation (NUPE C) is responsible for verification test. The fuel are subjected to post irradiation examination (PIE) and no unfavorable indications of fuel behavior have found both in NUPE C verification test and joint irradiation test by utilities and vendors. Burnup extension is an urgent task for LWR fuel in Japan in order to establish the domestic fuel cycle. It is conducted in joint efforts of industries, government and institutes. However, watching a situation of burnup extension in the world, we are not going ahead of other countries in the achievement of burnup extension. It is due to a conservative policy in the nuclear safety of the country. This is the reason why the burnup extension program in Japan is progressing 'slow and steady' As for the data obtained, no unfavorable indications of fuel behavior have found both in NUPE C verification test and joint irradiation test by utilities and vendors until now

  20. Development of Inverse Estimation Program of Burnup Histories for Nuclear Spent Fuel Based on ORIGEN-S

    International Nuclear Information System (INIS)

    The purpose of this work is to develop a computer program which can accurately estimate burnup histories of spent fuels based on the environmental sample measurements. The burnup histories of spent fuels include initial uranium enrichment, discharge burnup, cooling time after discharge, and nuclear reactor type in which the spent fuel was burnt. The methodologies employed in our program are based on the formulations developed by M. R. Scott1 but we developed a stable bi-section method to correct initial uranium enrichment and used a simplified algorithm without burnup correction. Also, ORIGEN-S2 rather than ORIGEN-23 was used in our program to improve the accuracies by using the new capabilities of burnup dependent cross section libraries of ORIGEN-S. Our program is applied to several benchmark problems including realistic Mihama-3 problems to test the accuracies. We developed a computer program to determine the burnup history such as initial uranium enrichment, burnup, cooling time, and reactor type by using the results of sample measurements as input. Our methodologies are based on the methodologies given in Ref. 1 but we devised a new stable bisection method for the correction of initial uranium enrichment and we used ORIGEN-S rather than ORIGEN-2 to utilize the new capabilities of ORIGEN-S such as burnup dependent cross sections which can be prepared by using SCALE6

  1. Incentives for the allowance of ''burnup credit'' in the design of spent nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    An analysis has been completed which indicates that the consideration of spent fuel histories ('burnup credit') in the criticality design of spent fuel shipping casks could result in significant public risk benefits and cost savings in the transport of spent nuclear fuel. Capacities of casks could be increased considerably in some cases. These capacity increases result in lower public and occupational exposures to ionizing radiation due to the reduced number of shipments necessary to transport a given amount of fuel. Additional safety benefits result from reduced non-radiological risks to both public and occupational sectors. In addition, economic benefits result from lower in-transit shipping costs, reduced transportation fleet capital costs, and fewer cask handling requirements at both shipping and receiving facilities

  2. Incentives for the allowance of burnup credit in the design of spent nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    An analysis has been completed which indicates that the consideration of spent fuel histories ('burnup credit') in the criticality design of spent fuel shipping casks could result in considerable public risk benefits and cost savings in the transport of spent nuclear fuel. Capacities of casks could be increased considerably in some cases. These capacity increases result in lower public and occupational exposures to ionizing radiation due to the reduced number of shipments necessary to transport a given amount of fuel. Additional safety benefits result from reduced non-radiological risks to both public and occupational sectors. In addition, economic benefits result from lower in-transit shipping costs, reduced transportation fleet capital costs, and fewer cask handling requirements at both shipping and receiving facilities

  3. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  4. Determination of nuclear fuel burnup by non-destructive gamma spectroscopy

    International Nuclear Information System (INIS)

    The determination of nuclear fuel burnup by the non-destructive gamma spectroscopy method is studied. A MTR (Materials Testing Reactor) -type fuel element is used in the measurement. The fuel element was removed from the reactor core in 1958 and, because of the long decay time, show only one peak in is gamma spectrum at 661.6 Kev. Corresponding to 137Cs. Measurements are made at 330 points of the element using a Nal detector and the final result revealed that the quantity of 235U consumed was 3.3 +- 0,8 milligram in the entire element. The effect of the migration of 137Cs in the element is neglected in view of the fact that it occurs only when the temperature is above 10000C, which is not the case in IEAR-1. (Author)

  5. Nuclear Energy Research Initiative. Development of a Stabilized Light Water Reactor Fuel Matrix for Extended Burnup

    International Nuclear Information System (INIS)

    The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble dopants in the UO2 matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus

  6. Analysis of high burnup spent nuclear fuel by ICP-MS

    International Nuclear Information System (INIS)

    Inductively coupled plasma mass spectrometry (ICP-MS) as the primary tool for determining concentrations of a suite of nuclides in samples excised from high-burnup spent nuclear fuel rods taken from light water nuclear reactors. The complete analysis included the determination of 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 137Cs, 143Nd, 145Nd, 148Nd, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 151Eu, 153Eu, 155Eu, 155Gd, 237Np, 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 242mAm, and 243Am. The isotopic composition of fissiogenic lanthanide elements was determined using high-performance liquid chromatography (HPLC) with ICP-MS detection. These analytical results allow the determination of fuel burn-up based on 148Nd, Pu, and U content, as well as provide input for storage and disposal criticality calculations. Results show that ICP-MS along with HPLC-ICP-MS are suitable of performing routine determinations of most of these nuclides, with an uncertainty of ±10% at the 95% confidence level. (author)

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

    International Nuclear Information System (INIS)

    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 keff 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. Determination of burnup balance for nuclear reactor fuel on the basis of γ-spectrometric determination of fission products

    International Nuclear Information System (INIS)

    Results are given of experimental investigations in one of the versions of the method for determination of the balance of nuclear fuel burnup process by means of the γ-spectrometry of fission products. In the version being considered a balance of the burnup process was determined on the base of 106Ru, 134Cs.Activity was measured by means of a γ-spectrometer with Ge counter. Investigations were done on the natural uranium metal fuel from the heavy-water moderated reactor of the first Czechoslovakian nuclear power plant A1 in Yaslovske Bohunice. Possibility was checked of determination of the fuel burnup depth as well as of the isotope ratio and content of plutonium. Results were compared with the control data which had been obtained on the base of the mass-spectrometry of U, Pu and Nd. The reasors for deviations were estimated in the cases when they were greater tan error in the control data

  9. Analysis of neodymium 148 in order to determin of nuclear fuel burnup

    International Nuclear Information System (INIS)

    To determine the degree of the nuclear fuel burnup experiments were conducted to introduce improvements in the mass-spectrometric study of neodymium-148 by the method of isotopic dilution with Nd-150 taken as a diluent. The separation of neodymium out of the mixture of the fission products and uranium was carried out in two stages. In the first stage a group of rare earth elements was isolated on the Vofatit SBV anionite in the mixture of nitric acid and methanol. The second stage involved the separation of the rare earth group on the Vofatite KPS cationite with the aid of the complexing agent of α-hydroxy-isobutyric acid. To identify the neodymium fraction, the traces of americium-241 were added at elution. The possibilities of the above analytical method are examplified by the isolation of neodymium out of the burned-up fuel of type EK-10. The isotopic ratios were determined by the spectroscopic method to the accuracy of +-1.2%. A highly enriched compound of neodymium-150 was used as a diluent. The factors are discussed affecting the degree of the burnup obtained by this method

  10. Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

    International Nuclear Information System (INIS)

    Highlights: ► Elastic constants of nuclear fuel were measured by acoustic microscopy. ► Our approach was in line with existing literature on non-irradiated material. ► Measurements on several samples of irradiated fuel (HBRP and N118) were performed. ► A decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. ► This trend is in good agreement with measurements conducted with indentation method. - Abstract: We report the measurement of elastic constants of non-irradiated UO2, SIMFUEL (simulated spent fuel: UO2 with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young’s modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.

  11. Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

    International Nuclear Information System (INIS)

    Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements' burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element's reported burnup or provide a burnup estimate for an element with an unknown burnup. (authors)

  12. A computer program for nuclear fuel burnup determination using gamma spectrometric methods

    International Nuclear Information System (INIS)

    In the end of its service life in the reactor, the fuel needs to be characterized for reasons relating both to safety and economy. The main investigations carried out are oriented towards verifying the fuel cladding integrity and determining the fissile content and the fuel burnup. A computer program for fast burnup evaluation was developed at the Post-Irradiation Examination Laboratory (PIEL) from INR Pitesti, the only laboratory of this kind in Romania. The input data consists, on one hand, of axial and radial gamma-scanning profiles (for the experimental evaluation of the number of nuclei of a given fission product - selected as burnup monitor - in the end of irradiation) and, on the other hand, of the history of irradiation (the time length and relative value of the neutron flux for each step of irradiation). Using the equation for the build-up and decay of the burnup monitor during irradiation the flux value is iteratively adjusted until the calculated number of nucleus is equal to the experimental one. Then the flux value is used in the equations of evolution of the fissile and fertile nuclei to determine the number of fissions and consequently the fuel burnup. The program was successfully used in the analysis of more then one hundred of TRIGA and CANDU-type fuel rods. An experimental result is reported in some details. (authors)

  13. Criticality safety evaluation for the direct disposal of used nuclear fuel. Preparation of data for burnup credit evaluation (Contract research)

    International Nuclear Information System (INIS)

    In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. In the conventional criticality safety evaluation of UNF where the fresh fuel composition is conservatively assumed, neutron multiplication factor is becoming overestimated as the fuel enrichment increases. The recent development of higher-enrichment fuel has therefore enhanced the benefit of the application of burnup credit. When applying the burnup credit to the criticality safety analysis of the disposed fuel system, the safe-side estimation of the reactivity is required taking into account the factors which affect the neutron multiplication factor of the burnt fuel system such as the nuclide composition uncertainties. In this report, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study. (author)

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

    International Nuclear Information System (INIS)

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

  15. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.)

  16. Evaluation of the elastic constants of the high burnup nuclear fuel (U1-y,Gdy)O2

    International Nuclear Information System (INIS)

    Nuclear fuels with burnable poisons are applications in the scope of the nuclear energy for electricity generation. The discharge burnups of this type of fuel are higher than the discharge burnups of pure UO2 fuels with same enrichment of the fissile element. In conditions of high burnups and moderate transients, the fuel rods can be subject to the pellet/cladding mechanical interactions (PCMI). Elastic constants of the fuel and cladding are used in the thermo-mechanical evaluation of the PCMI. The fuel elastic constants are usually expressed as a function of temperature, composition and porosity. Data survey and modeling of elastic constants for the (U1-y,Gdy)O2 fuel are developed in this paper. Due to small amount of the available measurements for the (U1-y,Gdy)O2, the modeling taken into account the data from UO2 and Gd2O3 measurements. Available measurements of the Al2O3 were also used in the model validation. The elastic constants of these materials have been fitted by an analytic equation, which depicts the whole range the dependences. (author)

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

    International Nuclear Information System (INIS)

    An increase of the discharge burn-up of UO2 nuclear fuels in the light water reactors results in the appearance of a change of microscopic structure, called HBS. Although well characterised experimentally, important points on the mechanisms of its formation remain to be cleared up. In order to answer these questions, a study of the contribution of the dislocation-type defects was conducted. In a first part, a calculation method of the stress field associated with periodic configurations of dislocations was developed. The method was applied to the cases of edge dislocation pile-up and wall, for which an explicit expression of the internal stress potential was obtained. Through the study of other examples of dislocation configurations, it was highlighted that this method also allows the calculation of any periodic dislocation configuration. In a second part, the evolution of interstitial-type dislocation loops was studied in UO2 fuel samples doped with 10% in mass of alpha emitters. The experimental loop size distributions were obtained for these samples stored during 4 and 7 years at room temperature. Kinetic equations are proposed in order to study the influence of the resolution process of interstitials from a loop back to the matrix due to an impact with the recoil atom 234U, as well as the coalescence of two interstitial loops that can diffuse by a volume mechanism. The application of the model shows that the two processes must be considered in the study of the evolution of radiation damage. (author)

  18. High-burnup fuel and the impact on fuel management

    International Nuclear Information System (INIS)

    Competition in the electric utility industry has forced utilities to reduce cost. For a nuclear utility, this means a reduction of both the nuclear fuel cost and the operating and maintenance cost. To this extent, utilities are pursuing longer cycles. To reduce the nuclear fuel cost, utilities are trying to reduce batch size while increasing cycle length. Yankee Atomic Electric Company has performed a number of fuel cycle studies to optimize both batch size and cycle length; however, certain burnup-related constraints are encountered. As a result of these circumstances, longer fuel cycles make it increasingly difficult to simultaneously meet the burnup-related fuel design constraints and the technical specification limits. Longer cycles require fuel assemblies to operate for longer times at relatively high power. If utilities continue to pursue longer cycles to help reduce nuclear fuel cost, changes may need to be made to existing fuel burnup limits

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

    International Nuclear Information System (INIS)

    Micrographs of uranium dioxide (UO2) corresponding to exposure times in reactor during 323, 953, 971, 1266 and 1642 full power days were investigated. The micrographs were converted into digital files isomorphous to cellular automata (CA) checkerboards. Such a representation of the fuel structure provides efficient tools for its dynamics simulation in terms of primary 'entities' imprinted in the micrographs. Besides, it also ensures a possibility of very effective micrograph processing by CA means. Interconnection between the description of fuel burn-up development and some exactly soluble models is ascertained. Evidences for existence of self-organization in the fuel at high burn-ups were established. The fractal dimension of microstructures is found to be an important characteristic describing the degree of radiation destructions

  20. Fuel cycle economical improvement by reaching high fuel burnup

    International Nuclear Information System (INIS)

    Improvements of fuel utilization in the light water reactors, burnup increase have led to a necessity to revise strategic approaches of the fuel cycle development. Different trends of the fuel cycle development are necessary to consider in accordance with the type of reactors used, the uranium market and other features that correspond to the nuclear and economic aspects of the fuel cycle. The fuel burnup step-by-step extension Program that successfully are being realized by the leading, firms - fuel manufacturers and the research centres allow to say that there are no serious technical obstacles for licensing in the near future of water cooling reactors fuel rod burnup (average) limit to 65-70 MWd/kgU and fuel assembly (average) limit to (60-65) MWd/kgU. The operating experience of Ukrainian NPPs with WWER-1000 is 130 reactor * years. At the beginning of 1999, a total quantity of the fuel FA discharged during all time of operation of 11 reactors was 5819 (110 fuel cycles). Economical improvement is reached by increase of fuel burn-up by using of some FA of 3 fuel cycles design in 4th fuel loading cycle. Fuel reliability is satisfactory. The further improvement of FA is necessary, that will allow to reduce the front-end fuel cycle cost (specific natural uranium expenditure), to reduce spent fuel amount and, respectively, the fuel cycle back end costs, and to increase burn-up of the fuel. (author)

  1. Effects of fuel burn-up and cooling periods on thermal responses in a repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    Flexible methods and codes have been developed to calculate thermal responses in a spent nuclear fuel repository located in granitic bedrock. The thermal resistance of the backfill material makes an essential contribution to the temperature at the canister surface. The backfill is however so thin compared to the rock masses of the repository that a stationary solution for the temperature drop across the backfill can be used. Combining analytical solutions of thermal diffusion in the rock and the stationary temperature following the released heat power across the backfill gives a very good description of the thermal behaviour of a repository. Temperatures at different points in a repository were calculated for encapsuled spent fuel with different cooling times (10, 20, 30 and 40 years) and having different burn-up values (33, 35 and 45 MWd/kgU). The amount of spent fuel in each canister was supposed to originate from 1.4 tons of fresh Uranium in the calculated examples. A more crucial variable in a thermal analysis of a repository configuration is however the heat power of canisters as a function of time and not the content of canisters. The study shows that the largest temperature rise at the boundary of a centrally located canister and buffer mass will be at most 85 K with assumed thermal data, when the heat power of 900 canisters does not exceed 1200-1400 W each at the time of emplacement depending on burn-up and cooling time. Assuming a higher value for the heat conductivity of the buffer material corresponding to that of the water saturated bentonite gives 15% higher limits respectively. The canisters were supposed to be placed in 15 rows 25 m apart from each other and 60 canisters in each row. The distance between canisters was taken to be 6 m

  2. High Burnup Fuel Performance and Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

    2007-03-15

    The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

  3. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  4. Non-destructive burn-up degree evaluation method for nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Makoto; Kumanomido, Hironori

    1998-01-06

    The present invention concerns a non-destructive burn-up degree evaluation method for spent fuels by a spontaneous neutron releasing rate method. Namely, an equation (1) is provided as: S = ({phi}/P)x(1-k) where {phi} is spontaneous neutron flux, P is the proportional coefficient, S is neutron releasing rate and k is neutron effective multiplication factor. S is further given by an equation (2): S = S4{sub 0}x(1+S2/S4{sub 0})xVxT where S2 is releasing rate from Cm242, S4{sub 0} is releasing rate from other nuclides, v is a void ratio of coolants and T is a time decaying effect, and the equations (1) and (2) are joined. P is determined by theoretical calculation, and S2/S4{sub 0} is determined based on a half decay characteristics of Cm242 to determine a correction amount. S4{sub 0} and V are determined as a correlational function of the burn-up degree: x based on burning calculation while using the Pu enrichment degree {epsilon}, Pu compositional ratio f, and concrete void ratio v. k is determined as a correlational function of v. A first appropriate value of x is obtained while having the burnup degree x{sup (0)} as an initial value. x is determined successively by repeating calculation based on modified k in this case. (I.S.)

  5. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    In fact, till now, burnup credit has not be applied in Ukrainian nuclear power for spent fuel management systems (storage and transport). However, application of advanced fuel at VVER reactors, arising spent fuel amounts, represent burnup credit as an important resource to decrease spent fuel management costs. The paper describes spent fuel management status in Ukraine from viewpoint of subcriticality assurance under spent fuel storage and transport. It also considers: 1. Regulation basis concerning subcriticality assurance, 2. Basic spent fuel and transport casks characteristics, 3. Possibilities and demands for burnup credit application at spent fuel management systems in Ukraine. (author)

  6. Implementation of burnup credit in PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    Implementation of burnup credit in spent fuel storage of LWR fuel at nuclear power plants is approved in Germany since the beginning of 2000. The burnup credit methods applied have to comply with the newly developed German criticality safety standard DIN 25471 passed in November 1999 and published in September 2000, cp. (orig.)

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

    International Nuclear Information System (INIS)

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

  8. Determination of Plutonium Contribution to the Total Burnup of a Spent Nuclear Fuel by Mass Spectrometric Measurements of Uranium and Ruthenium

    International Nuclear Information System (INIS)

    The U and Ru isotope patterns provide information on the real irradiation characteristics which are necessary for evaluating a fuel's performance in a reactor. A comparison of the Pu contribution values determined independently provides a promising way to check on the validity of the results. In order to check the consistency of the post-irradiation analysis results, correlations between the parameters of the irradiated nuclear fuels such as the concentration of the heavy elements and fission products, ratios of their isotopes and burnup were established. These correlations can be used to identify the reactor fuels and to estimate the burnup and Pu production. A new approach was carried out with Ru isotopic ratio for the determination of Pu contribution to the total burnup of a spent nuclear fuel from a power reactor. The principle of this approach was based on the use of the difference in the fission yield ratios of the Ru fission products involved for the three main fissionable nuclides such as 235U, 239Pu, and 241Pu. In this work, to determine the contribution of Pu to the total burnup of the fuel, the following two independent methods have been applied: by measuring the isotope ratios of the stable Ru fission products 101Ru/104Ru, and by determining the total burnup by Nd-148 method and subtracting partial burnup, which is determined form the measured values of U isotope ratios

  9. Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

    International Nuclear Information System (INIS)

    The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit

  10. Calculation study of TNPS spent fuel pool using burnup credit

    International Nuclear Information System (INIS)

    Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks, the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied. The MONK9A code was used to analyze keff, of different enrichment fuels at different burnups. A reference loading curve was proposed in accordance with the system keff's changing with the burnup of different initially enriched nuclear fuels. The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

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

  12. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

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

    International Nuclear Information System (INIS)

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

  14. Implementation of burnup and control rod credit for storage of spent nuclear fuel in Ukraine

    International Nuclear Information System (INIS)

    Preliminary analysis of the regulations in force in Ukraine concerning nuclear safety of spent nuclear fuel management systems shows that some regulatory requirements in force are too conservative in view of current international practice. The extent of conservatism can be determined and reduced, if necessary, only using calculated studies for analyzing the criticality of spent nuclear fuel management systems. Such activity is consistent with the requirements posed by state-of-the-art production requirements. However, this can be only based on improving our level of understanding the processes occurring in nuclear dangerous systems and improving our capabilities as regards accuracy, correctness, and reliability in numerical modeling these processes. This work was intended to demonstrate that the excessive conservatism laid previously into the requirements on nuclear safety in Ukraine due to insufficient development of means for modeling processes in nuclear fuel can be considerably decreased through using more real modeling fuel systems. If such modeling is performed with the use of state-of-the-art software and computers, based on more complete understanding the processes in fuel systems, then removal of the excessive conservatism does will not reduce the safety of nuclear dangerous systems. (author)

  15. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  16. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    International Nuclear Information System (INIS)

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

  17. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  18. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    International Nuclear Information System (INIS)

    Highlights: • The burnup of irradiated AGR-1 TRISO fuel was analyzed using gamma spectrometry. • The burnup of irradiated AGR-1 TRISO fuel was also analyzed using mass spectrometry. • Agreement between experimental results and neutron physics simulations was excellent. - Abstract: AGR-1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR-1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non-destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR-1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs-137 activity and the other based on the ratio of Cs-134 and Cs-137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA (fissions per initial heavy metal atom) for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can be determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP-MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma

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

    International Nuclear Information System (INIS)

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

  20. Re-evaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

    International Nuclear Information System (INIS)

    Highlights: → The specifications required for the analyses of the destructive assay data taken from irradiated fuel in Ohi-1 and Ohi-2 PWRs were documented in this paper. → These data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. → These destructive assay data are suitable for the benchmarking of the burnup calculation code systems. - Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.

  1. 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.

  2. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

    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

  3. Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety

  4. Experimental investigation of burnup credit for safe transport, storage, and disposal of spent nuclear fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Berry, Donald T.; Harms, Gary A.; Ford, John T.; Walker, Sharon Ann; Helmick, Paul H.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  5. Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Harms, Gary A.; Helmick, Paul H.; Ford, John T.; Walker, Sharon A.; Berry, Donald T.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  6. Detailed Burnup Calculations for Testing Nuclear Data

    Science.gov (United States)

    Leszczynski, F.

    2005-05-01

    -section data for burnup calculations, using some of the main available evaluated nuclear data files (ENDF-B-VI-Rel.8, JEFF-3.0, JENDL-3.3), on an isotope-by-isotope basis as much as possible. The selected experimental burnup benchmarks are reference cases for LWR and HWR reactors, with analysis of isotopic composition as a function of burnup. For LWR (H2O-moderated uranium oxide lattices) four benchmarks are included: ATM-104 NEA Burnup credit criticality benchmark; Yankee-Rowe Core V; H.B.Robinson Unit 2 and Turkey Point Unit 3. For HWR (D2O-moderated uranium oxide cluster lattices), three benchmarks were selected: NPD-19-rod Fuel Clusters; Pickering-28-rod Fuel Clusters; and Bruce-37-rod Fuel Clusters. The isotopes with experimental concentration data included in these benchmarks are: Se-79, Sr90, Tc99, Ru106, Sn126, Sb125,1129, Cs133-137, Nd143, 145, Sm149-150, 152, Eul53-155, U234-235, 238, Np237, Pu238-242, Am241-243, and Cm242-248. Results and analysis of differences between calculated and measured absolute and/or relative concentrations of these isotopes for the seven benchmarks are included in this work.

  7. Burnup effects of MOX fuel pincells in PWR - OECD/NEA burnup credit benchmark analysis -

    International Nuclear Information System (INIS)

    The burnup effects were analyzed for various cases of MOX fuel pincells of fresh and irradiated fuels by using the HELIOS, MCNP-4/B, CRX and CDP computer codes. The investigated parameters were burnup, cooling time and combinations of nuclides in the fuel region. The fuel compositions for each case were provided by BNFL (British Nuclear Fuel Limited) as a part of the problem specification so that the results could be focused on the calculation of the neutron multiplication factor. The results of the analysis show that the largest saving effect of the neutron multiplication factor due to burnup credit is 30 %. This is mainly due to the consideration of actinides and fission products in the criticality analysis

  8. Extended burnup: fuel development and performance

    International Nuclear Information System (INIS)

    Fuel Performance for the B and W 15 x 15 (Mark B) and 17 x 17 (Mark C) fuel assembly designs is examined on a plant by plant basis. An extensive data base of fuel assembly and rod bow measurements and tests which demonstrate that these phenomena should not limit the high burnup capability of B and W fuel is presented. Post-irradiation measurements to date for fuel rod and assembly growth show that these phenomena are behaving as predicted and can be adequately evaluated and designed for in high burnup fuel assemblies. Clad creep and ductility data as a function of burnup for B and W fuel is presented with emphasis on their effects on our high burnup targets. Finally, fission gas release and waterside corrosion measurements results are presented

  9. Application of a burnup verification meter to actinide-only burnup credit for spent PWR fuel

    International Nuclear Information System (INIS)

    A measurement system to verify reactor records for burnup of spent fuel at pressurized water reactors (PWR) has been developed by Sandia National Laboratories and tested at US nuclear utility sites. The system makes use of the Fork detector designed at Los Alamos National Laboratory for the safeguards program of the International Atomic Energy Agency. A single-point measurement of the neutrons and gamma- rays emitted from a PWR assembly is made at the center plane of the assembly while it is partially raised from its rack in the spent fuel pool. The objective of the measurements is to determine the variation in burnup assignments among a group of assemblies, and to identify anomalous assemblies that might adversely affect nuclear criticality safety. The measurements also provide an internal consistency check for reactor records of cooling time and initial enrichment. The burnup verification system has been proposed for qualifying spent fuel assemblies for loading into containers designed using burnup credit techniques. The system is incorporated in the US Department of Energy's.''Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages'' [DOE/RW 19951

  10. Nuclear fuel behaviour modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The Technical Committee Meeting (TCM) included separate sessions on the specific topics of fuel thermal performance and fission product retention. On thermal performance, it is apparent that the capability exists to measure conductivity in high burnup fuel either by out-of-pile measurement or by instrumentation of test reactor rods. State-of-the-art modelling codes contain models for the conductivity degradation process, and hence adequate predictions of fuel temperature are achievable. Concerning fission product release, it is clear that many groups around the world are actively investigating the subject, with experimental and modelling programmes being pursued. However, a general consensus on the exact mechanisms of gas release and related gas bubble swelling has yet to emerge, even at medium burnup levels. Fission gas phenomena, not only the release to open volumes, but the whole sequence of processes taking place prior to this, need to be modelled in any modern fuel performance code. The presence of gaseous fission products may generate rapid fuel swelling during power transients, and this can cause PCI and rod failure. At high burnups, the quantity of released gases could give rise to pressures exceeding the safe limits. Modelling of pellet-cladding interaction (PCI) effects during transient operation is also an active area of study for many groups. In some situations a purely empirical approach to failure modelling can be justified, while for other applications a more detailed mechanistic approach is required. Another aspect of cladding modelling which was featured at the TCM concerned corrosion and hydriding. Although this issue can be the main life-limiting factor on fuel duty, it is apparent that modelling methods, and the experimental measurement techniques that underpin them, are adequate. A session was included on MOX fuel modelling. Substantial programmes of work, especially by the MOX vendors, appear to be underway to bring the level of understanding

  11. Evolution of the nuclear fuel mechanical properties at high burn-up at extensive European experimental program

    International Nuclear Information System (INIS)

    Computer codes have been developed in order to simulate the nuclear fuel rod mechanical behaviour, and therefore compare the evolution of the main parameters against a certain number of safety criteria, for reactor class 1 and class 2 operating conditions. Concerning the fuel material (UO2, MOX or UO2 with additives) the mechanical properties have been determined on non irradiated samples. One can expect that these properties evolve with burn-up, due to the transmutation, the evolution of the oxygen potential, the accumulation of fission defects and in some case the material restructuring (High Burn-up Structure). In order to provide the fuel thermo-mechanical calculation with more accurate mechanical properties, a large experimental project has been launched since several years by CEA, EDF and ITU ; furthermore a recent collaboration has been started with Studsvik, Sweden, concerning the possibility of performing in-pile creep measurements. The program is indeed organised in three folds, which can be described as follows: 1 - Acquisition of mechanical properties on non irradiated materials (UO2, UO2+Gd, UO2+Cr, MOX) with axial creep tests, three points bending tests up to 1700 oC, acoustic measurements at room temperature, instrumented micro-indentation tests and Vickers test from room temperature to 1200 oC. This allowed the development of a mechanical behaviour law available for these materials in the non-irradiated state. 2 - Acquisition of mechanical properties on irradiated materials in hot cells, using a micro-indentation machine developed especially in TUI, Vickers tests with the same machine, and a focused acoustic technique developed by the LAIN laboratory in the Montpellier University (France). The target is to define the evolution of the elastic properties, of the yield stress and of the thermal creep properties. 3 - More recently with the Studsvik Laboratories, the design of a specific rig for in-pile indentation has been completed. This aims at

  12. Burnup and plutonium distribution of WWER-440 fuel pin at extended burnup

    International Nuclear Information System (INIS)

    The formation of rim region in LWR UO2 based nuclear fuel at high burnup is a common observation. This region has very high porosity due to excessive gas release. Such a region is also characterized by a significantly high plutonium concentration and high local burnup compared to the internal fuel region. Spatial distribution of these parameters has been incorporated with fuel behavior and performance analysis codes by using mostly empirical relations. Variation of these parameters depends on the neutron flux as well as neutron energy spectrum. Detailed neutronics analysis is necessary for the accurate prediction of these parameters. This study is performed by MCNP4B Monte Carlo code for the calculation of local neutron flux, ORIGEN2 for burnup and depletion calculations, and MONTEBURNS for coupling these codes. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell. Fuel pin is divided into a number of radial segments. A relatively small mesh size is used at the region near the surface to reveal the rim effect. The variation of plutonium and local burnup are obtained for high burnup. Results are compared with existing experimental observations for WWER-440 fuel and other theoretical predictions

  13. Burnup credit considerations in dry spent-fuel storage licensing

    International Nuclear Information System (INIS)

    Burnup credit has been allowed in reactor basin spent-fuel storage at pressurized water reactors for a number of years. However, such storage occurs under strict administrative, procedural, and design controls. In recent years, dry spent-fuel storage cask vendors have expressed interest in designing cask fuel baskets with allowance for burnup credit. At last year's American Nuclear Society Winter Meeting, an ad hoc session was organized and authorized on burnup credit for dry storage and transportation casks. It has become clear that some utilities are interested in burnup credit for dry storage designs. Given this, the US Nuclear Regulatory Commission (NRC) staff is examining the technical issues involved in allowing burnup credit. Analytical work focused on the development of branch technical positions for determination of burnup credit for dry spent-fuel storage technology designs has begun. Procedural and administrative issues will be examined, based on licensing experience, and will also be the subject of branch technical positions. At an appropriate time, preparation of regulatory guides will be considered

  14. PWR fuel performance and burnup extension programme in Japan

    International Nuclear Information System (INIS)

    Since the first PWR nuclear power plant Mihama Unit 1 initiated commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts on improving the technology of PWRs. The results can already be seen by the significantly improved performance of the PWR plants now in operation. Mitsubishi Heavy Industries, Ltd supplied the nuclear fuel assemblies, which now amount to almost 5000. Although some trouble with fuel was experienced in the beginning, the progressive efforts made to improve the fuel design and manufacturing technology have resulted in the superior performance of Mitsubishi fuels. Since fuel of current design should comply with the limitation set in Japan for a maximum discharged fuel assembly average burnup of less than 39,000 MW·d/t, the maximum burnup is now around 37,000 MW·d/t. However, an increase in this burnup limitation has been strongly requested by Japanese utilities in order to make nuclear power more economic and thus more competitive with other power generation methods. A summary is given of the design improvements made on Mitsubishi fuel, as well as demonstration programmes of current design fuel to prove its superior reliability and to prepare the database for a future extension of burnup. (author)

  15. Burnup credit implementation in spent fuel management

    International Nuclear Information System (INIS)

    The criticality safety analysis of spent fuel management systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. The concept of allowing reactivity credit for spent fuel offers economic incentives. Burnup Credit (BUC) could reduce mass limitation during dissolution of highly enriched PWR assemblies at the La Hague reprocessing plant. Furthermore, accounting for burnup credit enables the operator to avoid the use of Gd soluble poison in the dissolver for MOX assemblies. Analyses performed by DOE and its contractors have indicated that using BUC to maximize spent fuel transportation cask capacities is a justifiable concept that would result in public risk benefits and cost savings while fully maintaining criticality safety margins. In order to allow for Fission Products and Actinides in Criticality-Safety analyses, an extensive BUC experimental programme has been developed in France in the framework of the CEA-COGEMA collaboration. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Independent measurement systems, e.g. gamma spectrum detection systems, are needed to perform a true independent measurement of assembly burnup, without reliance on reactor records, using the gamma emission signatures fission products (mainly Cesium isotopes). (author)

  16. Characterization of the non-uniqueness of used nuclear fuel burnup signatures through a Mesh-Adaptive Direct Search

    Science.gov (United States)

    Skutnik, Steven E.; Davis, David R.

    2016-05-01

    The use of passive gamma and neutron signatures from fission indicators is a common means of estimating used fuel burnup, enrichment, and cooling time. However, while characteristic fission product signatures such as 134Cs, 137Cs, 154Eu, and others are generally reliable estimators for used fuel burnup within the context where the assembly initial enrichment and the discharge time are known, in the absence of initial enrichment and/or cooling time information (such as when applying NDA measurements in a safeguards/verification context), these fission product indicators no longer yield a unique solution for assembly enrichment, burnup, and cooling time after discharge. Through the use of a new Mesh-Adaptive Direct Search (MADS) algorithm, it is possible to directly probe the shape of this "degeneracy space" characteristic of individual nuclides (and combinations thereof), both as a function of constrained parameters (such as the assembly irradiation history) and unconstrained parameters (e.g., the cooling time before measurement and the measurement precision for particular indicator nuclides). In doing so, this affords the identification of potential means of narrowing the uncertainty space of potential assembly enrichment, burnup, and cooling time combinations, thereby bounding estimates of assembly plutonium content. In particular, combinations of gamma-emitting nuclides with distinct half-lives (e.g., 134Cs with 137Cs and 154Eu) in conjunction with gross neutron counting (via 244Cm) are able to reasonably constrain the degeneracy space of possible solutions to a space small enough to perform useful discrimination and verification of fuel assemblies based on their irradiation history.

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

    International Nuclear Information System (INIS)

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

  18. The Design Method for the ATR High Burnup MOX Fuel

    International Nuclear Information System (INIS)

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has developed the advanced thermal reactor (ATR). PNC is demonstrating MOX fuel utilization in a prototype of ATR, Fugen (165 MWe), in which 638 MOX fuel assemblies have been loaded without a failure since 1979. PNC is developing the high burn-up MOX fuel for the ATR to contribute to MOX fuels for thermal reactors. The statistical design evaluation method that included the MOX fuel rod performance evaluation code 'FEMAXI-ATR' was developed for the ATR high bum-up MOX fuel rod; it was verified that the integrity of the fuel could be maintained over the whole irradiation period

  19. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  20. NFCSim: A Dynamic Fuel Burnup and Fuel Cycle Simulation Tool

    International Nuclear Information System (INIS)

    NFCSim is an event-driven, time-dependent simulation code modeling the flow of materials through the nuclear fuel cycle. NFCSim tracks mass flow at the level of discrete reactor fuel charges/discharges and logs the history of nuclear material as it progresses through a detailed series of processes and facilities, generating life-cycle material balances for any number of reactors. NFCSim is an ideal tool for analysis - of the economics, sustainability, or proliferation resistance - of nonequilibrium, interacting, or evolving reactor fleets. The software couples with a criticality and burnup engine, LACE (Los Alamos Criticality Engine). LACE implements a piecewise-linear, reactor-specific reactivity model for its criticality calculations. This model constructs fluence-dependent reactivity traces for any facility; it is designed to address nuclear economies in which either a steady state is never obtained or is a poor approximation. LACE operates in transient and equilibrium fuel management regimes at the refueling batch level, derives reactor- and cycle-dependent initial fuel compositions, and invokes ORIGEN2.x to carry out burnup calculations

  1. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

  2. 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.

  3. Nuclear fuel activities in Canada

    International Nuclear Information System (INIS)

    Nuclear fuel activities in Canada are considered in the presentation on the following directions: Canadian utility fuel performance; CANDU owner's group fuel programs; AECL advanced fuel program (high burnup fuel behaviour and development); Pu dispositioning (MOX) activities. 1 tab

  4. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

  5. Burnup credit in nuclear waste transport: An economic analysis

    International Nuclear Information System (INIS)

    The US DOE is responsible for transporting nuclear spent fuel from commercial reactors to monitored retrievable storage (MRS) facilities and/or to repositories. Current plans call for approximately 110,000 metric tons uranium (MTU) to be transported over approximately 40 years beginning in 1998. Because of the large volume of spent fuel to be transported, new generations of spent fuel transportation casks are being planned. These casks will embody the latest technology and will be designated to accommodate the spent fuel in a way that maximizes the overall efficiency of the cask. In planning for the new generation of transport casks, the DOE is investigating the possibility of tailoring the cask design for the extent to which spent fuel has been used in the reactors, or, for spent fuel burnup. Granting design credit for burnup would allow one to fabricate casks with relatively larger capacities than would be possible otherwise. The remainder of the paper discusses the economic implications of using burnup credit in cask design, discusses the approach used in analyzing the economics of burnup credit, describes the results of the analysis, and offers some conclusions about the economic value of the burnup credit option

  6. Burnup monitoring of VVER-440 spent fuel assemblies

    International Nuclear Information System (INIS)

    This paper reports on the results of the experiments performed on spent VVER-440 fuel assemblies at the Paks Nuclear Power Plant (NPP), Hungary. The fuel assemblies submerged in the service pit were examined by high-resolution gamma spectrometry (HRGS). The assemblies were moved to the front of a collimator tube built in the concrete wall of the pit in the reactor block at the NPP, and lifted down and up under water for scanning by the refueling machine. The HPGe detector was placed behind the collimator in an outside staircase. The measurements involved scanning of the assemblies along their length of all the 6 sides, at 5-12 measurement positions side by side. Axial and azimuthal burnup profiles were taken in this way. Assembly groups for measurements were selected according to their burnup (10–50 GWd/tU) and special positions (e. g. control assembly, neighbour of control assembly). Burnup differences were well observable between assembly sides looking towards the center of the core and opposite directions. Also, burnup profiles were different for control assemblies and normal (working) fuel assemblies. The ratio of the measured activities of Cs-134 and Cs-137 was evaluated by relative efficiency (intrinsic) calibration. Measurement uncertainty is around 3 %. Taking into account irradiation history and cooling time (i. e.the time elapsed since the discharge of the assembly out of the core), the activity ratio Cs-134/Cs-137 shows good correlation with the declared burnup.

  7. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

    Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of 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, followed by criticality calculations to assess the value of keff for 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 used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel

  8. Compressive creep of simulated burnup fuel

    International Nuclear Information System (INIS)

    In order to study the nitride fuel mechanical properties, we measured the compressive steady state creep rates of uranium mononitride (UN) and UN containing neodymium which was simulated burnup fuel. The stress exponent n'' and the apparent activation energy ''Q'' of the creep rate were determined in the range of 27.5 ≤ σ ≤ 200.0 MPa and 950 ≤ T ≤ 1500 degC. (author)

  9. Light a CANDLE. An innovative burnup strategy of nuclear reactors

    International Nuclear Information System (INIS)

    CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)

  10. Trade-off and optimization of fuel cycle costs in high burnup fuel management schemes

    International Nuclear Information System (INIS)

    Evaluations of the fuel cycle costs of nuclear reactors normally consider uranium ore procurement, conversion to hex, enrichment, fuel fabrication, transport at the front-end and back-end costs such as spent fuel interim storage, transport and direct disposal/reprocessing. The methods for carrying out such evaluation are firmly established and generally show a clear incentive to increase discharge burnups in order to benefit from improved fuel cycle economics. This paper challenges the conventional approach to fuel cycle economics, arguing that there are additional considerations that should legitimately be included in fuel cycle cost calculations. An illustrative calculation o fuel cycle costs for high burnup cycles with allowances for such additional factors shows that fuel cycle costs are a minimum at around 55 GWd/t discharge burnup. (authors)

  11. Reactivity effects of nonuniform axial burnup distributions on spent fuel

    International Nuclear Information System (INIS)

    When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup

  12. Reactivity effects of nonuniform axial burnup distributions on spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Leary, R.W. II; Parish, T.A. [Texas A & M Univ., College Station, TX (United States)

    1995-12-01

    When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup.

  13. Irradiation test for verification of PWR 48 GWd/t high burnup fuel

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) has conducted the irradiation test for verification of the high burnup fuel performance under the sponsorship of the Ministry of Economy, Trade and Industry. (NUPEC-HB Project) As for PWR, the fuel burnup is extended by two steps. The Step I fuel (maximum fuel assembly discharge burnup: 48 GWd/t), has been utilized since 1989. And now, the preparation for the regular utilization of Step II fuel (maximum fuel assembly discharge burnup: 55 GWd/t), is being conducted. The results of pre- and post-irradiation tests on the Step I fuel irradiated in the Takahama-3 of Kansai Electric Power Co., Inc., were analyzed and evaluated. The irradiation performance of fuel rod, pellet, cladding and fuel assembly showed no remarkable difference compared with that of other published paper. Consequently the reliability and integrity of the Step I fuel was verified. (author)

  14. Activity ratio measurement and burnup analysis for high burnup PWR fuels

    International Nuclear Information System (INIS)

    Applying burnup credit to spent fuel transportation and storage system is beneficial. To take burnup credit to criticality safety design for a spent fuel transportation cask and storage rack, the burnup of target fuel assembly based on core management data must be confirmed by experimental methods. Activity ratio method, in which measured the ratio of the activity of a nuclide to that of another, is one of the ways to confirm burnup history. However, there is no public data of gamma-ray spectrum from high burnup fuels and validation of depletion calculation codes is not sufficient in the evaluation of the burnup or nuclide inventories. In this study, applicability evaluation of activity ratio method was carried out for high burnup fuel samples taken from PWR lead use assembly. In the gamma-ray measurement experiments, energy spectrum was taken in the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA), and 134Cs/137Cs and 154Eu/137Cs activity ratio were obtained. With the MVP-BURN code, the activity ratios were calculated by depletion calculation tracing the operation history. As a result, 134Cs/137Cs and 154Eu/137Cs activity ratios for UO2 fuel samples show good agreements and the activity ratio method has good applicability to high burnup fuels. 154Eu/134Cs activity ratio for Gd2O3+UO2 fuels also shows good agreements between calculation results and experimental results as well as the activity ratio for UO2 fuels. It also becomes clear that it is necessary to pay attention to not only burnup but also axial burnup distribution history when confirming the burnup of UO2+Gd2O3 fuel with 134Cs/137Cs activity ratios. (author)

  15. Economic viability to BeO-UO2 fuel burnup extension

    International Nuclear Information System (INIS)

    This paper presents the quantitative analysis results of research on the burnup effect on the nuclear fuel cycle cost of BeO-UO2 fuel. As a result of this analysis, if the burnup is 60 MWD/kg, which is the limit under South Korean regulations, the nuclear fuel cycle cost is 4.47 mills/kWh at 4.8wt% of Be content for the BeO-UO2 fuel. It is, however, reduced to 3.70 mills/kWh at 5.4wt% of Be content if the burnup is 75MWD/kg. Therefore, it seems very advantageous, in terms of the economic aspect, to develop BeO-UO2 fuel, which does not have any technical problem with its safety and is a high burnup and long life cycle nuclear fuel

  16. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  17. Improving burnup performance of fast sodium cooled reactor by utilizing thorium based fuels

    International Nuclear Information System (INIS)

    To study the improvement of fuel burnup for fast reactors, thorium based fuels are investigated. In order to ensure the projected expansion of nuclear power is achieved in conjunction with reduced risk of nuclear weapons proliferation, new conventional sources of fuel will have to be made available. Thorium fuel cycles have many incentives such as the reduction of plutonium generation and consumption of LWR actinides, the provision of high performance burnup, and the conservation of 235U resources. This work examined the burnup reactivity loss and depletion analysis of thorium versus uranium based metal fuels. When compared the thorium based metallic fuel outperformed uranium based fuel with respect to higher actinide burnup and higher depletion rate of plutonium isotopes. (authors)

  18. Burnup determination of water reactor fuel

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency in consultation with the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The meeting was hosted by the Commission of the European Communities, at the Transuranium Research Laboratory, Joint Research Centre Karlsruhe, in the Federal Republic of Germany. This subject was dealt with for the first time by the IAEA. It was found to correspond adequately to this type of Specialist Meeting and to be suitable at a moment when the extension of burnup constitutes a major technical and economical issue in fuel technology. It was stressed that analysis of highly burnt fuels, mixed oxides and burnable absorber bearing fuels required extension of the experimental data base, to comply with the increasing demand for an improved fuel management, including better qualification of reactor physics codes. Twenty-seven participants from eleven countries plus two international organizations attended the Meeting. Twelve papers were given during three technical sessions, followed by a panel discussion which allowed to formulate the conclusions of the meeting and recommendations to the Agency. In addition, participants were invited to give an outline of their national programmes, related to Burnup Determination of Water Reactor Fuel. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

  19. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  20. Burnup credit implementation in WWER spent fuel management systems: Status and future aspects

    International Nuclear Information System (INIS)

    This paper describes the motivation for possible burnup credit implementation in WWER spent fuel management systems in Bulgaria. The activities being done are described, namely: the development and verification of a 3D few-group diffusion burnup model; the application of the KORIGEN code for evaluation of WWER fuel nuclear inventory during reactor core lifetime and after spent fuel discharge; using the SCALE modular system (PC Version 4.1) for criticality safety analyses of spent fuel storage facilities. Future plans involving such important tasks as validation and verification of computer systems and libraries for WWER burnup credit analysis are shown. (author)

  1. PIE Results and New Techniques Applied for 55GWd/t High Burnup Fuel of PWR

    International Nuclear Information System (INIS)

    Post-irradiation examinations (PIE) for 55GWd/t high burnup fuel which had been irradiated at a domestic PWR plant was conducted at the fuel hot laboratory of the Nuclear Development Corporation (NDC). In this PIE, such new techniques as the clamping for axial tensile test and the pellets density measurement method for high burnup fuels were used in addition to existing techniques to confirm the integrity of 55GWd/t high burnup fuel. The superiority of improved corrosion-resistant claddings over currently used current Zircaloy-4 claddings in terms of corrosion-resistance was also confirmed. This paper describes the PIE results and the advanced PIE techniques. (author)

  2. Determination of the accuracy of utility spent fuel burnup records. Interim report

    International Nuclear Information System (INIS)

    In order to develop a NRC-licensable burnup credit methodology, the pedigree and uncertainty of commercial spent nuclear fuel assembly burnup records needs to be established. Typically the assembly average burnup for each assembly is maintained in the plant records. It is anticipated that the repository for the disposal of spent fuel will utilize burnup credit and will require knowledge of the uncertainty of reactor burnup records. The uncertainty of the assembly average burnup record depends on the uncertainty of the method used to develop the record. Such records are generally based on core neutronic analysis coupled with analysis of in-core power detector data. This report evaluates the uncertainties in the burnup of fuel assemblies utilizing in-core measurements and core neutronic calculations for a Westinghouse PWR. To quantify the uncertainty, three cycles of in-core movable detector data were used. The data represents a first cycle of operation, a transition cycle and a low leakage cycle. These three cycles of data provide a true test of the uncertainty methodology. Three separate sets of results were used to characterize the burnup uncertainty of the fuel assemblies. The first set of results compared the measured and calculated reaction rates in instrumented assemblies and determined the uncertainty in the reaction rates. The second set of results determined the uncertainty in relative assembly power for both the instrumented and un-instrumented assemblies. The third set of results determined the burnup uncertainty of the discharged fuel in each cycle

  3. 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.

  4. Assessment of the use of extended burnup fuel in light water power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

    1988-02-01

    This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

  5. Determination of research reactor fuel burnup

    International Nuclear Information System (INIS)

    This report was prepared by a Consultants Group which met during 12-15 June 1989 at the Jozef Stefan Institute, Yugoslavia, and during 11-13 July 1990 at the IAEA Headquarters in Vienna, Austria, with subsequent contributions from the Consultants. The report is intended to provide information to research reactor operators and managers on the different, most commonly used methods of determining research reactor fuel burnup: 1) reactor physics calculations, 2) measurement of reactivity effects, and 3) gamma ray spectrometry. The advantages and disadvantages of each method are discussed. References are provided to assist the reactor operator planning to establish a programme for burnup determination of fuel. Destructive techniques are not included since such techniques are expensive, time consuming, and not normally performed by the reactor operators. In this report, TRIGA fuel elements are used in most examples to describe the methods. The same techniques however can be used for research reactors which use different types of fuel elements. 22 refs, 13 figs, 2 tabs

  6. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  7. Evaluation of a high burnup spent fuel regarding the regulations for a spent fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ik Sung; Yang, Young Sik; Bang, Je Geon; Kim, Dae Ho; Kim, Sun Ki; Song, Keun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    All nuclear plants have storage pools for spent fuel. These pools are typically 40 or more feet deep. In many countries, the spent fuels are stored under water. The water serves 2 purposes: 1) It serves as a shield to reduce the radiation levels. 2) It cools the fuel assemblies that continue to produce heat (called decay heat). But Korean nuclear plant expects the storage capacity to reach its limit by the year 2016. So, the research for the spent fuel dry storage facilities is necessary. The purpose of this study was to overview the regulatory basis for spent fuel dry storage and to evaluate its applicability for high burnup spent fuel.

  8. Development of base technology for high burnup PWR fuel improvement Volume 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Eun; Lee, Sang Hee; Bae, Seong Man [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Chung, Jin Gon; Chung, Sun Kyo; Kim, Sun Du [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Kim, Jae Won; Chung, Sun Kyo; Kim, Sun Du [Korea Nuclear Fuel Development Inst., Seoul (Korea, Republic of)

    1995-12-31

    Development of base technology for high burnup nuclear fuel -Development of UO{sub 2} pellet manufacturing technology -Improvement of fuel rod performance code -Improvement of plenum spring design -Study on the mechanical characteristics of fuel cladding -Organization of fuel failure mechanism Establishment of next stage R and D program (author). 226 refs., 100 figs.

  9. Experimental programmes related to high burnup fuel

    International Nuclear Information System (INIS)

    The experimental programmes undertaken at IGCAR with regard to high burn-up fuels fall under the following categories: a) studies on fuel behaviour, b) development of extractants for aqueous reprocessing and c) development of non-aqueous reprocessing techniques. An experimental programme to measure the carbon potential in U/Pu-FP-C systems by methane-hydrogen gas equilibration technique has been initiated at IGCAR in order to understand the evolution of fuel and fission product phases in carbide fuel at high burn-up. The carbon potentials in U-Mo-C system have been measured by this technique. The free energies and enthalpies of formation of LaC2, NdC2 and SmC2 have been measured by measuring the vapor pressures of CO over the region Ln2O3-LnC2-C during the carbothermic reduction of Ln2O3 by C. The decontamination from fission products achieved in fuel reprocessing depends strongly on the actinide loading of the extractant phase. Tri-n-butyl phosphate (TBP), presently used as the extractant, does not allow high loadings due to its propensity for third phase formation in the extraction of Pu(IV). A detailed study of the allowable Pu loadings in TBP and other extractants has been undertaken in IGCAR, the results of which are presented in this paper. The paper also describes the status of our programme to develop a non-aqueous route for the reprocessing of fast reactor fuels. (author)

  10. Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias

    International Nuclear Information System (INIS)

    Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)

  11. Investigation of research and development subjects for very high burnup fuel. Development of fuel cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Nagase, Fumihisa; Suzuki, Masahide; Furuta, Teruo; Suzuki, Yasufumi; Hayashi, Kimio; Amano, Hidetoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1993-05-01

    Plutonium use as well as burnup extension of UO{sub 2} fuel is an important subject for the strategy of utilization of the nuclear energy in LWRs. A higher burnup is favorable to MOX fuel in economic respect and for effective use of plutonium. Therefore, the concept of a `very high burnup` aiming at the maximum bundle burnup of 100GWd/t has been proposed assuming use of MOX fuel. The authors have investigated research and development subjects for the fuel pellet and the cladding material to be developed. The present report shows the results on the cladding material. In order to achieve a very high burnup, development of the cladding material with higher corrosion and radiation resistance compared with Zircaloy is necessary. In this report, zirconium based alloy, stainless steel, nickel and titanium based alloys, ceramics, etc. were reviewed considering water corrosion resistance, thermal and mechanical properties, radiation effects, etc. Furthermore, capability of these materials as the fuel cladding was discussed focusing on water side corrosion and radiation effect on mechanical properties. As a result, candidate materials at present and the required research tasks were shown with issues for the development. (author) 66 refs.

  12. Actinide-only burnup credit for spent fuel transport

    International Nuclear Information System (INIS)

    A conservative methodology is described that would allow taking credit for burn up in the criticality safety analysis of spent nuclear fuel packages. Requirements for its implementation include isotopic and criticality validation, generation of package loading criteria using limiting parameters, and assembly burn up verification by measurement. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentrations with burnup. No credit for fission product neutron absorbers is taken. Analyses are included regarding the methodology's financial benefits and conservative margin. It is estimated that the proposed actinide-only burnup credit methodology would save 20% of the transport costs. Nevertheless, the methodology includes a substantial margin. Conservatism due to the isotopic correction factors, limiting modelling parameters, limiting axial profiles and exclusion of the fission products ranges from 10 to 25% k. (author)

  13. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

    Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1/2 truck cask, which has been relicensed to carry pressurized water reactor (PWR) fuel with 56,000 MWd/ton U burnup at 450 days of cooling time. New cask designs should consider the effects of high burnup for future use, even though it is not commercially desirable to include currently unneeded capability. In conclusion, the increased heat and gamma radiation of high-burnup fuels can be accommodated by additional cooling time, but the increased neutron radiation source cannot be accommodated unless the balance of neutron and gamma contributions to the overall dose rate is properly chosen in the initial cask design. Criticality control of high-burnup fuels is possible with heavily poisoned baskets, but burnup credit in licensing is a much more direct means of demonstrating criticality safety

  14. Current Status of Burnup Evaluation for Test Fuel at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Park, Seung Jae; Shin, Yoon Taeg; Choo, Kee Nam; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For the research reactor, 8 mini plate fuels were irradiation-tested during 4 irradiation cycles. 2 more irradiation capsules were fabricated for additional test of plate type fuel. Also fission Mo target for the performance verification and the demonstration of Mo-99 extraction process will be irradiated at HANARO. It is important to evaluate the burnup history of test fuel. The burnup of test fuel has been calculated using HANARO Fuel Management System (HANAFMS). Although it is proper to evaluate the burnup of HANARO fuel, it is difficult to accurately calculate the burnup of test fuel due to the limitation of HANAFMS model. Therefore, the improvement of burnup evaluation for the recent irradiated test fuel is conducted and reported in this paper. To evaluate the burnup of test fuel, HANAFMS has been used; however, HANAFMS model is not proper to apply plate type fuel. Therefore, MCNP burned core model was developed for HAMP-1 burnup calculation. Throughout the comparison of fuel assembly power, MCNP burned core model showed the good agreement with HANAFMS.

  15. Taking burnup credit for interim storage and transportation system for BWR fuels

    International Nuclear Information System (INIS)

    In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)

  16. Fuel burnup characteristics for the NRU research reactor

    International Nuclear Information System (INIS)

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U3Si, consisting of particles of U3Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  17. Fuel burnup characteristics for the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C., E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U{sub 3}Si, consisting of particles of U{sub 3}Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  18. BNFL assessment of methods of attaining high burnup MOX fuel

    International Nuclear Information System (INIS)

    It is clear that in order to maintain competitiveness with UO2 fuel, the burnups achievable in MOX fuel must be enhanced beyond the levels attainable today. There are two aspects which require attention when studying methods of increased burnups - cladding integrity and fuel performance. Current irradiation experience indicates that one of the main performance issues for MOX fuel is fission gas retention. MOX, with its lower thermal conductivity, runs at higher temperatures than UO2 fuel; this can result in enhanced fission gas release. This paper explores methods of effectively reducing gas release and thereby improving MOX burnup potential. (author)

  19. Radionuclide Release from High Burnup Fuel

    International Nuclear Information System (INIS)

    In this paper we investigate the production, evolution and release of radioactive fission products in a light water reactor. The production of the nuclides is determined by the neutronics, their evolution in the fuel by local temperature and by the fuel microstructure and the rate of release is governed by the scenario and the properties of the microstructure where the nuclides reside. The problem combines fields of reactor physics, fuel behaviour analysis and accident analysis. Radionuclide evolution during fuel reactor life is also important for determination of instant release fraction of final repository analysis. The source term problem is investigated by literature study and simulations with reactor physics code Serpent as well as fuel performance code ENIGMA. The capabilities of severe accident management codes MELCOR and ASTEC for describing high burnup structure effects are reviewed. As the problem is multidisciplinary in nature the transfer of information between the codes is studied. While the combining of the different fields as they currently are is challenging, there are some possibilities to synergy. Using reactor physics tools capable of spatial discretization is necessary for determining the HBS inventory. Fuel performance studies can provide insight how the HBS should be modelled in severe accident codes, however the end effect is probably very small considering the energetic nature of the postulated accidents in these scenarios. Nuclide release in severe accidents is affected by fuel oxidation, which is not taken into account by ANSI/ANS-5.4 but could be important in some cases, and as such, following the example of severe accident models would benefit the development of fuel performance code models. (author)

  20. WWER fuel behaviour and characteristics at high burnup

    International Nuclear Information System (INIS)

    The increase of fuel burnup in fuel rods is a task that provides a considerable cost reduction of WWER fuel cycle in case of its solution. Investigations on fuel and cladding behaviour and change in fuel characteristics under irradiation are carried out in the Russian Federation for standard and as well as for experimental fuel rods to validate the reliable and safe operation of the fuel rods at high burnups. The paper presents the results of examinations on cracking, dimensional, structural and density changes of fuel pellets as well as the results of examination on corrosion and mechanical properties of WWER-440 and WWER-1000 fuel rod claddings. (author)

  1. Implementation of burnup credit in spent fuel management systems

    International Nuclear Information System (INIS)

    Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate safety margin. The motivation for using burnup credit in criticality safety applications is based on economic considerations and additional benefits contributing to public health and safety and resource conservation. Interest in the implementation of burnup credit has been shown by many countries. In 1997, the International Atomic Energy Agency (IAEA) started a task to monitor the implementation of burnup credit in spent fuel management systems, to provide a forum to exchange information, to discuss the matter and to gather and disseminate information on the status of national practices of burnup credit implementation in the Member States. The task addresses current and future aspects of burnup credit. This task was continued during the following years. (author)

  2. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    International Nuclear Information System (INIS)

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided

  3. Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

  4. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    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)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  5. Introduction of new flasks for high burnup spent fuel

    International Nuclear Information System (INIS)

    New flasks have been designed to transport the high burnup spent fuels now becoming available from the world's nuclear power stations. Two versions have been designed: Excellox 6 for 5 metre PWR fuels and Excellox 7 with increased neutron shielding for 4.5 metre PWR and BWR fuels arising in Japan. The designs of these flasks have been finalised; Excellox 6 has been approved and validated as a Type B(U)F package and the first two have been manufactured and are now in routine service, with a third at an advanced stage of manufacture. The Excellox 7 design is ready for manufacture when service requirements for it have been settled. An account is given of the final adjustments to the design in the course of manufacture, the main steps and tests in the manufacturing process and the commissioning tests at the reprocessing and reactor sites. The entry of the flasks into service is reviewed. (author)

  6. New Strategies for Licensing the Storage and Transportation of High Burn-up Spent Nuclear Fuel in the United States - 12546

    International Nuclear Information System (INIS)

    An alternative approach may be needed to the licensing of high-burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates. Packaging strategies and regulations should be developed to reduce the potential for requiring fuel to be repackaged unnecessarily. This would lessen the chance of accidents and mishaps during loading and unloading of casks, and decrease dose to workers. A packaging approach that shifts the safety basis from reliance upon the fuel condition to reliance upon an inner canister could eliminate or lessen the need for repackaging. In addition, the condition of canisters can be more readily monitored and inspected than the condition of fuel cladding. Canisters can also be repaired and/or replaced when deemed necessary. In contrast, once a fuel assembly is loaded into a canister and placed in a storage overpack, there is little opportunity to monitor its condition or take mitigating measures if cladding degradation is suspected or proven to occur. (authors)

  7. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  8. Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

    International Nuclear Information System (INIS)

    This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

  9. Raman micro-spectroscopy of UOX and MOX spent nuclear fuel characterization and oxidation resistance of the high burn-up structure

    Science.gov (United States)

    Jegou, C.; Gennisson, M.; Peuget, S.; Desgranges, L.; Guimbretière, G.; Magnin, M.; Talip, Z.; Simon, P.

    2015-03-01

    Raman micro-spectroscopy was applied to study the structure and oxidation resistance of UO2 (burnup 60 GWd/tHM) and MOX (burnup 47 GWd/tHM) irradiated fuels. The Raman technique, adapted to working under extreme conditions, enabled structural information to be obtained at the cubic micrometer scale in various zones of interest within irradiated fuel (central and zones like the Rim for UOX60, and the plutonium-enriched agglomerates for MOX47 characterized by a high burn-up structure), and the study of their oxidation resistance. As regards the structural information after irradiation, the spectra obtained make up a set of data consistent with the systematic presence of the T2g band characteristic of the fluorite structure, and of a triplet band located between 500 and 700 cm-1. The existence of this triplet can be attributed to the presence of defects originating in changes to the fuel chemistry occurring in the reactor (presence of fission products) and to the accumulation of irradiation damage. As concerns the oxidation resistance of the different zones of interest, Raman spectroscopy results confirmed the good stability of the restructured zones (plutonium-enriched agglomerates and Rim) rich in fission products compared to the non-restructured UO2 grains. A greater structural stability was noticed in the case of high plutonium content agglomerates, as this element favors the maintenance of the fluorite structure.

  10. BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies

    International Nuclear Information System (INIS)

    BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inventory as a function of time and power density. The code allows for multiple fueled regions to be analyzed. The companion code, RELOAD, can be used to shuffle fueled regions or reload regions with fresh fuel. BURNCAL can be used to study the reactivity effects and isotopic inventory as a function of time for a nuclear reactor system. Neutron transmutation, fission, and radioactive decay are included in the modeling of the production and removal terms for each isotope of interest. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. Fueled and un-fueled regions, such as cladding and moderator, can be analyzed simultaneously. The nuclides analyzed are limited only by the neutron cross section availability in the MCNP cross-section library. BURNCAL is unique in comparison to other burnup codes in that it does not use the calculated neutron flux as input to other computer codes to generate the nuclide mixture for the next time step. Instead, BURNCAL directly uses the neutron absorption tally/reaction information generated by MCNP for each nuclide of interest to determine the nuclide inventory for that region. This allows for the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed

  11. Research on irradiation behavior of superhigh burnup fuel

    International Nuclear Information System (INIS)

    In Japan Atomic Energy Research Institute, the special team for LWR future technology development project was organized in Tokai Research Establishment from October, 1991 to the end of fiscal year 1993. Due to the delay of the introduction of fast reactors, LWRs are expected to be used for considerably long period also in 21st century, therefore, it aimed at the further advancement of LWRs, and as one of its embodiments, the concept of superhigh burnup fuel was investigated. The superhigh burnup fuel aims at the attainment of 100 GWd/t burnup, and it succeeded the achievement of the conceptual design study on 'superlong life LWRs'. It is generally recognized that the development of the new material that substitutes for zircaloy is indispensable for superhigh burnup fuel. The concept of superhigh burnup core and the specification of fuel, the research and development of superhigh burnup fuel, the research on the irradiation behavior and irradiation damage of fuel and the damage by ion irradiation, and the method and the results of the irradiation experiment using a tandem accelerator are reported. (K.I.)

  12. Research on irradiation behavior of superhigh burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-03-01

    In Japan Atomic Energy Research Institute, the special team for LWR future technology development project was organized in Tokai Research Establishment from October, 1991 to the end of fiscal year 1993. Due to the delay of the introduction of fast reactors, LWRs are expected to be used for considerably long period also in 21st century, therefore, it aimed at the further advancement of LWRs, and as one of its embodiments, the concept of superhigh burnup fuel was investigated. The superhigh burnup fuel aims at the attainment of 100 GWd/t burnup, and it succeeded the achievement of the conceptual design study on `superlong life LWRs`. It is generally recognized that the development of the new material that substitutes for zircaloy is indispensable for superhigh burnup fuel. The concept of superhigh burnup core and the specification of fuel, the research and development of superhigh burnup fuel, the research on the irradiation behavior and irradiation damage of fuel and the damage by ion irradiation, and the method and the results of the irradiation experiment using a tandem accelerator are reported. (K.I.).

  13. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  14. Impact of High Burnup Uranium Oxide and Mixed Uranium-Plutonium Oxide Water Reactor Fuel on Spent Fuel Management

    International Nuclear Information System (INIS)

    There is increasing worldwide use of uranium oxide (UOX) nuclear fuel with higher enrichments and burnups as the reliability of UOX fuel increases and the economics of moving to higher burnup fuel improves. Burnup extension affects all important stages of the nuclear fuel cycle and thus concerns the entire nuclear industry. There are many aspects of switching to higher burnup UOX or MOX fuels, such as reliability, safety, and economics, that decision makers need to take into account. The potential physical changes to the fuel rods and assemblies will affect the operation of the components of the back end of the fuel cycle. The objective of this report is to provide information on the impacts on spent fuel management to those countries operating LWRs and HWRs with zirconium alloy clad UOX fuels who are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties of higher burnup UOX and MOX are potentially significant. Examples of the differences in properties between lower burnup UOX and higher burnup UOX and MOX include: higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties. Higher burnup UOX or MOX usage affects all spent fuel management components, such as wet and dry storage, transportation, reprocessing, re-fabricated fuel and final disposal. This report briefly reviews the current fuel characteristics of UOX and MOX and the potential for characteristic changes with increased burnup. In addition, the components of the back end of the fuel cycle are discussed. Evaluation of these effects on the back end of the fuel cycle was based on the particular fuel behaviour, regulatory, safety, sustainability, or operational issues that might be impacted by the increased burnup or switch to MOX. Other than a brief mention, an economic evaluation of the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

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

    International Nuclear Information System (INIS)

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

  17. Determination of Fission Gas Inclusion Pressures in High Burnup Nuclear Fuel using Laser Ablation ICP-MS combined with SEM/EPMA and Optical Microscopy

    International Nuclear Information System (INIS)

    In approximately 20% of all fissions at least one of the fission products is gaseous. These are mainly xenon and krypton isotopes contributing up to 90% by the xenon isotopes. Upon reaching a burn-up of 60 - 75 GWd/tHM a so called High Burnup Structure (HBS) is formed in the cooler rim of the fuel. In this region a depletion of the noble fission gases (FG) in the matrix and an enrichment of FG in μm-sized pores can be observed. Recent calculations show that in these pores the pressure at room temperature can be as large as 30 MPa. The knowledge of the FG pressure in pores is important to understand the high burn-up fuel behavior under accident conditions (i.e. RIA or LOCA). With analytical methods routinely used for the characterization of solid samples, i.e. Electron Probe Micro Analysis (EPMA), Secondary Ion Mass Spectrometry (SIMS), the quantification of gaseous inclusions is very difficult to almost impossible. The combination of a laser ablation system (LA) with an inductively coupled plasma mass spectrometer (ICP-MS) offers a powerful tool for quantification of the gaseous pore inventory. This method offers the advantages of high spatial resolution with laser spot sizes down to 10 μm and low detection limits. By coupling with scanning electron microscopy (SEM) for the pore size distribution, EPMA for the FG inventory in the fuel matrix and optical microscopy for the LA-crater sizes, the pressures in the pores and porosity was calculated. As a first application of this calibration technique for gases, measurements were performed on pressurized water reactor (PWR) fuel with a rod average of 105 GWd/tHM to determine the local FG pressure distribution. (authors)

  18. Physical mechanism analysis of burnup actinide composition in light water reactor MOX fuel and its application to uncertainty evaluation

    International Nuclear Information System (INIS)

    Highlights: • We discuss physical mechanisms for burnup actinide compositions in LWR’s MOX fuel. • Mechanisms of 244Cm and 238Pu productions are analyzed in detail with sensitivity. • We can evaluate the indirect effect on actinide productions by nuclear reactions. • Burnup sensitivity is applied to uncertainty evaluation of nuclide production. • Actinides can be categorized into patterns according to a burnup sensitivity trend. - Abstract: In designing radioactive waste management and decommissioning facilities, understanding the physical mechanisms for burnup actinide composition is indispensable to satisfy requirements for its validity and reliability. Therefore, the uncertainty associated with physical quantities, such as nuclear data, needs to be quantitatively analyzed. The present paper illustrates an analysis methodology to investigate the physical mechanisms of burnup actinide composition with nuclear-data sensitivity based on the generalized depletion perturbation theory. The target in this paper is the MOX fuel of the light water reactor. We start with the discussion of the basic physical mechanisms for burnup actinide compositions using the reaction-rate flow chart on the burnup chain. After that, the physical mechanisms of the productions of Cm-244 and Pu-238 are analyzed in detail with burnup sensitivity calculation. Conclusively, we can identify the source of actinide productions and evaluate the indirect influence of the nuclear reactions if the physical mechanisms of burnup actinide composition are analyzed using the reaction-rate flow chart on the burnup chain and burnup sensitivity calculation. Finally, we demonstrate the usefulness of the burnup sensitivity coefficients in an application to determine the priority of accuracy improvement in nuclear data in combination with the covariance of the nuclear data. In addition, the target actinides and reactions are categorized into patterns according to a sensitivity trend

  19. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

    The examinations carried out on the FRAGEMA fuel of EDF reactors show its good behavior in service. The results of research and development programs developed by EDF, FGA and the CEA show that this fuel can be irradiated up to a high burnup, and allow to point out the axies of research to improve still the performance of the product in a more and more soliciting environment (increase of power and burnup coupled with load following). Among the solutions considered, there are the design and fabrication adjustments (geometry, initial pressurization), more fundamental changes concerning fuel cans and fuel pellets, which need still research and development programs

  20. Analysis of burnup credit on spent fuel storage

    International Nuclear Information System (INIS)

    Chemical analyses were carried out on high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins. Measured data of the composition of nuclides from 234U to 242Pu were used for evaluation of ORIGEN-2/82 code. Criticality calculations were executed for the casks which were being designed to store 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for (1) axial and horizontal profiles of burnup, and void history (BWR), (2) operational histories such as control rod insertion history, BPR insertion history and others, and (3) calculational accuracy of ORIGEN-2/82 code on the composition of nuclides. Present evaluation shows that introduction of burnup credit has a substantial merit in criticality safety analysis of the cask, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for present reactivity bias evaluation and showed a possibility of simplifying the reactivity bias evaluation in burnup credit. Finally, adapting procedures of burnup credit such as the burnup meter were evaluated. (author)

  1. Criticality evaluation of high density spent fuel storge rack under normal condition using burnup credit

    International Nuclear Information System (INIS)

    The high density spent fuel storage rack Boraflex was known to experience changes of its physical property and to dissolve under exposure to radiation in an aqueous environment for long period of time. In this study, the criticality evaluation for spent fuel storage rack of Ulchin Unit 2 under normal condition was performed assuming complete loss of 10B from the Boraflex and applying burnup credit. Criticality evaluation code KENO-V.a. from SCALE4.4 system was benchmarked against critical experiments to obtain the calculation bias and bias uncertainties. The manufacturing tolerances of nuclear fuel and storage rack and their reactivity uncertainties were derived, as well. Considering those bias and uncertainties of calculation, the criticality of spent fuel storage under normal condition was conservatively evaluated. The criticality evaluation result using burnup credit can be presented as a spent fuel loading curve that indicates the acceptable burnup domain in spent fuel storage pool. The spent fuels with various initial enrichments and discharge fuel burnup can be safely accommodated in the storage without taking any boron credit from Boraflex, provided the combination falls within the acceptable domain in the loading curve. The spent fuel with initial enrichment of 5.0w/o was evaluated to meet the subcritical safety if its burnup is over 43.0GWD/MTU. The criticality evaluation result also showed that spent fuels with the initial enrichment less than 1.6w/o were able to be stored in the storage pool regardless of their burnup. Conclusively, in the Region 2 of the spent fuel storage pool, the maximum keff , considering all uncertainties, was calculated as 0.94818

  2. A complete NUHOMS registered solution for storage and transport of high burnup spent fuel

    International Nuclear Information System (INIS)

    The discharge burnups of spent fuel from nuclear power plants keep increasing with plants discharging or planning to discharge fuel with burnups in excess of 60,000 MWD/MTU. Due to limited capacity of spent fuel pools, transfer of older cooler spent fuel from fuel pool to dry storage, and very limited options for transport of spent fuel, there is a critical need for dry storage of high burnup, higher heat load spent fuel so that plants could maintain their full core offload reserve capability. A typical NUHOMS registered solution for dry spent fuel storage is shown in the Figure 1. Transnuclear, Inc. offers two advanced NUHOMS registered solutions for the storage and transportation of high burnup fuel. One includes the NUHOMS registered 24PTH system for plants with 90.7 Metric Ton (MT) crane capacity; the other offers the higher capacity NUHOMS registered 32PTH system for higher crane capacity. These systems include NUHOMS registered - 24PTH and -32PTH Transportable Canisters stored in a concrete storage overpack (HSM-H). These canisters are designed to meet all the requirements of both storage and transport regulations. They are designed to be transported off-site either directly from the spent fuel pool or from the storage overpack in a suitable transport cask

  3. Testing of experimental fuel elements for VVER-1000 reactors in MR to high fuel burnup

    International Nuclear Information System (INIS)

    Pressurized water reactors are given a commanding role in the development program for the nation's nuclear power industry. Considerable operating experience has been gained with VVER-1000 reactors. As of the start of 1990, 17 units with VVER-1000 reactors were in operation in this country and abroad. The first loadings were designed for a 2-year run with average fuel burnup of 28.5 MW-day/kg. The rod-type fuel elements used in the reactors displayed high serviceability and reliability (leakage does not exceed 0.02%). Operating experience and the results of computational and experimental work have made it possible to substantiate the possibility of switching them to a 3-year run. The fifth unit of the Novovoronezh Atomic Power Plant was the first to be switched to a 3-year run, as of 1984. The average fuel burnup achieved after three fuel cycles was 42.6 MW-day/kg. All units with VVER-1000 reactors are now being switched to a 3-year run with an average burnup of more than 40 MW-day/kg for the unloaded fuel

  4. Final LDRD report : nanoscale mechanisms in advanced aging of materials during storage of spent %22high burnup%22 nuclear fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Clark, Blythe G.; Rajasekhara, Shreyas; Enos, David George; Dingreville, Remi Philippe Michel; Doyle, Barney Lee; Hattar, Khalid Mikhiel; Weiner, Ruth F.

    2013-09-01

    We present the results of a three-year LDRD project focused on understanding microstructural evolution and related property changes in Zr-based nuclear cladding materials towards the development of high fidelity predictive simulations for long term dry storage. Experiments and modeling efforts have focused on the effects of hydride formation and accumulation of irradiation defects. Key results include: determination of the influence of composition and defect structures on hydride formation; measurement of the electrochemical property differences between hydride and parent material for understanding and predicting corrosion resistance; in situ environmental transmission electron microscope observation of hydride formation; development of a predictive simulation for mechanical property changes as a function of irradiation dose; novel test method development for microtensile testing of ionirradiated material to simulate the effect of neutron irradiation on mechanical properties; and successful demonstration of an Idaho National Labs-based sample preparation and shipping method for subsequent Sandia-based analysis of post-reactor cladding.

  5. Behaviour of fuel rods of the second generation at high burnup WWER-440 fuel cycles. Aspects for attainment of burnup 70 MWd/kgU

    International Nuclear Information System (INIS)

    In this report an analysis of WWER-440 fuel of the second generation supplied by Russian JSC TVEL for high burnup fuel cycle is presented. The certificated code START-3 is applied to modeling of fuel rod operation parameters. Reliability of high-burnup fuel on the base of 5-6 year operation is demonstrated. Special attention is paid to aspects for attainment of burnup 70 MWd/kgU, including experimental and fuel modeling support and fuel operation experience

  6. Development of high burnup fuel data-base

    International Nuclear Information System (INIS)

    Development of high burnup fuel data base (HBDB) was studied, which stores various performance data of high burnup fuels using a personal computer. Data items of the data base and storing and display methods of time-depending data such as power history were studied. It was shown that compound systems of a personal computer and an engineering work station have capacity for constructing the data base with much efficiency and small cost. And comparison of data items between the data base and the EPRI fuel base FPDB was discussed. (author)

  7. Technical and economic limits to fuel burnup extension. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    For many years, the increase of efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase of fuel burnup leads to a reduction in the volume of spent fuel discharged to longer fuel cycles in the reactor, which means bigger availability and capacity factors. After having increased the authorized burnup in plants, developing new alloys capable of resisting high burnup, and having accumulated data on fuel evolution with burnup, it has become necessary to establish the limitations which could be imposed by the physical evolution of the fuel, influencing fuel management, neutron properties, reprocessing or, more generally, the management of waste and irradiated fuels. It is also necessary to verify whether the benefits of lower electricity costs would not be offset by an increase in fuel management costs. The main questions are: Are technical and economic limits to the increasing of fuel burnup in parallel? Can we envisage nowadays the hardest limitation in some of these areas? Which are the main points to be solved from the technical point of view? Is this effort worthwhile considering the economy of the cycle? To which extent? For these reasons, the IAEA, following a recommendation by the International Working Group on Fuel Performance and Technology, held a Technical Committee Meeting on Technical and Economic Limits to Fuel Burnup Extension. The purpose of this meeting was to provide an international forum to review the evolution of fuel properties at increased burnup in order to estimate the limitations both from a physical and an economic point of view. The meeting was therefore divided into two parts. The first part, focusing on technical limits, was devoted to the improvement of the fuel element, such as fission gas release (FGR), RIM effect, cladding, etc. and the fabrication, core management, spent fuel and reprocessing. Eighteen related papers were presented which

  8. Investigation of research and development subjects for very high burnup fuel

    International Nuclear Information System (INIS)

    Plutonium use as well as burnup extension of UO2 fuel is an important subject for the strategy of utilization of the nuclear energy in LWRs. A higher burnup is favorable to MOX fuel in economic respect and for effective use of plutonium. Therefore, the concept of a 'very high burnup' aiming at the maximum bundle burnup of 100GWd/t has been proposed assuming use of MOX fuel. The authors have investigated research and development subjects for the fuel pellet and the cladding material to be developed. The present report shows the results on the cladding material. In order to achieve a very high burnup, development of the cladding material with higher corrosion and radiation resistance compared with Zircaloy is necessary. In this report, zirconium based alloy, stainless steel, nickel and titanium based alloys, ceramics, etc. were reviewed considering water corrosion resistance, thermal and mechanical properties, radiation effects, etc. Furthermore, capability of these materials as the fuel cladding was discussed focusing on water side corrosion and radiation effect on mechanical properties. As a result, candidate materials at present and the required research tasks were shown with issues for the development. (author) 66 refs

  9. Transient behaviour of high burnup fuel

    International Nuclear Information System (INIS)

    The main subjects of the meeting were the discussion of regulatory background, integral tests and analysis, plant calculations, separate-effect test and analysis, concerning high burnup phenomena during RIA accidents in reactors, especially LWR, BWR and PWR type reactors. 32 papers were abstracted and indexed individually for the INIS database. (R.P.)

  10. Economics of VVER Fuel Cycles Leading to High Discharge Burnup

    International Nuclear Information System (INIS)

    Economic characteristics of equilibrium VVER fuel cycles leading to high discharge burnup are investigated by supposing two scenarios named optimistic and pessimistic. The optimistic and pessimistic terms are used in the sense whether the high burnup fuel cycles are economically advantageous or the increasing enrichment cost can increase the specific fuel cycle cost above a certain discharge burnup value. Therefore in case of the optimistic scenario, maximum fabrication and back end costs and minimum enrichment and raw uranium costs were applied, while in case of the pessimistic scenario vice-versa. The applied costs are detailed in Table 1. Table1 Cost data of the two different scenarios. Concerning the transport and storage during the front end fuel cycle, it was assumed that application of burnable poison solves the criticality problems caused by the increased enrichment. By using the advantage of the burnup credit, the subcriticality of the spent fuel storage and transport devices can also be proved. Large reserve in the biological shielding is supposed. According to the above argumentation, fixed cost of the front and back end fuel cycle was used in the calculations, except the enrichment, but a 700 $/pin extra fabrication cost of the burnable poison was taken into account. Instead of fixed batch fraction, fixed cycle length was assumed which is advantageous for maximizing the discharge burnup and for minimizing the burnable poison extra cost but disadvantageous concerning the availability factor, which is constant in the given calculations. Beside the economic characteristics, the feasibility of the cycles are investigated from the point of view of the most important safety related parameters like reactivity coefficients and shut down margin. The figure below shows the burnup dependent fuel cycle cost for the above two scenarios. (author)

  11. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Wang Tienko E-mail: tkw@faculty.nthu.edu.tw; Peir Jinnjer

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products {sup 97}Zr/{sup 97}Nb, {sup 132}I, and {sup 140}La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, {sup 235}U burn-up values can be deduced by iterative calculations. The complication caused by {sup 239}Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products {sup 137}Cs, {sup 134}Cs/{sup 137}Cs ratio and {sup 106}Ru/{sup 137}Cs ratio.

  12. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    International Nuclear Information System (INIS)

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio

  13. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry.

    Science.gov (United States)

    Wang, T K; Peir, J J

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

  14. ORIGEN computer code use in non-destructive analysis of irradiated fuel elements for burn-up determination

    International Nuclear Information System (INIS)

    An iterative method for burn-up determination in the non-destructive analysis of irradiated fuel elements using the ORIGEN computer code is presented. On the bases of data obtained from ORIGEN code the calibration coefficient for the neutron flux is determined as a function of one fission product activity while the burn-up is determined as a function of the calibration coefficient for a given irradiation history. These functions are used for determining the burn-up of nuclear fuel elements measured by gamma-scanning. The method is tested for fuel elements irradiated in a TRIGA reactor facility. (Author)

  15. In-reactor thermo-mechanical measurements on LWR fuel rods in the high burnup range

    International Nuclear Information System (INIS)

    The extension of fuel burn-up beyond previously accepted levels is currently being applied in varying degrees throughout the nuclear industry, with the aim of improving fuel economics and reducing the spent fuel volume. So it is necessary that the current fuel knowledge base should be extended. Modifications of fuel rod/assembly concepts, together with fuel management schemes, should be gradually implemented so that the operation of power reactors becomes even more reliable and flexible than it is today. Extrapolation to extended burn-up levels does not cause concern but will have to be made in steps, in order to demonstrate expected performance trends. The fuel testing programmes at the OECD Halden Reactor Project have over the years significantly contributed to the understanding of LWR fuel behaviour in the high burn-up range. A broad range of versatile and integrated in-reactor test rigs and high pressure loops have been developed which allow simulations of LWR irradiation conditions, comparative testing of alternative fuel rod designs and use of test segments pre-irradiated in power reactors. A number of in-core instruments and experimental techniques have been developed for detailed investigations of various aspects related to the thermal behaviour, fission product release and mechanical response of high burn-up LWR fuel rods, under a variety of operating conditions. The paper reviews recent measurements in the area of burnup-dependent steady-state and transient thermal behaviour of fuel rods, intermixing of fission and helium filler gases in the pellet cladding gap, fission gas release kinetics under changing heat loads and power excursions (burst release) and dimensional changes of fuel rods subjected to cyclic load changes. (author). 14 refs, 12 figs

  16. Burn-up measurement of irradiated rock-like fuels

    International Nuclear Information System (INIS)

    In order to obtain burn-up data of plutonium rock-like (ROX) fuels irradiated at JRR-3M in JAERI, destructive chemical analysis of zirconia or thoria system ROX fuels was performed after development of a new dissolution method. The dissolution method and procedure have been established using simulated ROX fuel, which is applicable to the hot-cell handling. Specimens for destructive chemical analysis were obtained by applying the present method to irradiated ROX fuels in a hot-cell. Isotopic ratios of neodymium and plutonium were determined by mass-spectrometry using the isotope dilution procedure. Burn-up of the irradiated ROX fuels was calculated by the 148Nd procedure using measured data. The burn-ups of thoria and zirconia system fuels that irradiated same location in the capsule showed almost same values. For the ROX fuel containing thorium, 233U was also determined by the same techniques in order to evaluate the effect of burn-up of thorium. As the result, it was found that the fission of 233U was below 1% of total fission number and could be negligible. In addition, americium and curium were determined by alpha-spectrometry. These data, together with isotopic ratio of plutonium, are important data to analyze the irradiation behavior of plutonium. (author)

  17. Prediction of fission gas pressure from high burnup oxide fuel

    International Nuclear Information System (INIS)

    The ELESIM fuel performance code incorporates a fundamentally based treatment of the relevant physical processes affecting fission gas release. The fission gas release model treats fission gas diffusion, formation and subsequent interlinkage of intergranular bubbles, grain boundary storage of gas, grain growth and fuel swelling. The latter case considers the contributions of thermal expansion, densification, solid fission products, and gas bubbles. The effect of porosity on fuel thermal conductivity is taken into account. Previously we showed predictions of the gas release model agreed well with measured values for oxide fuel with burnups to about 300 MW.h/kg U. The applicability of the model to high burnup fuel is examined using examples from the literature. The fission gas release range considered is about 1-100% for burnups to 1000 MW.h/kg U in thermal reactor fuel and 2400 MW.h/kg U in fast reactor fuel. Predicted and measured releases are shown to be in good agreement, suggesting that the fundamental model is correct. In some models, empirical correction factors are required at high burnup to achieve agreement between predicted and measured release values; no such factor is required in ELESIM. (auth)

  18. Fuel cycle cost considerations of increased discharge burnups

    International Nuclear Information System (INIS)

    Evaluations are presented that indicate the attainment of increased discharge burnups in light water reactors will depend on economic factors particular to individual operators. In addition to pure resource conserving effects and assuming continued reliable fuel performance, a substantial economic incentive must exist to justify the longer operating times necessary to achieve higher burnups. Whether such incentive will exist or not will depend on relative price levels of all fuel cycle cost components, utility operating practices, and resolution of uncertainties associated with the back-end of the fuel cycle. It is concluded that implementation of increased burnups will continue at a graduated pace similar to past experience, rather than finding universal acceptance of particular increased levels at any particular time

  19. Mechanical Property Evaluation of High Burnup PHWR Fuel Clads

    International Nuclear Information System (INIS)

    Assurance of clad integrity is of vital importance for the safe and reliable extension of fuel burnup. In order to study the effect of extended burnup of 15,000 MW∙d/tU on the performance of Pressurised Heavy Water Reactor (PHWR) fuel bundles of 19-element design, a couple of bundles were irradiated in Indian PHWR. The tensile property of irradiated cladding from one such bundle was evaluated using the ring tension test method. Using a similar method, claddings of mixed oxide (MOX) fuel elements irradiated in the pressurized water loop (PWL) of CIRUS to a burnup of 10,000 MW∙d/THM were tested. The tests were carried out both at ambient temperature and at 300°C. The paper will describe the test procedure, results generated and discuss the findings. (author)

  20. A burnup credit calculation methodology for PWR spent fuel transportation

    International Nuclear Information System (INIS)

    A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)

  1. Summary of high burnup fuel issues and NRC`s plan of action

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.O.

    1997-01-01

    For the past two years the Office of Nuclear Regulatory Research has concentrated mostly on the so-called reactivity-initiated accidents -- the RIAs -- in this session of the Water Reactor Safety Information Meeting, but this year there is a more varied agenda. RIAs are, of course, not the only events of interest for reactor safety that are affected by extended burnup operation. Their has now been enough time to consider a range of technical issues that arise at high burnup, and a list of such issues being addressed in their research program is given here. (1) High burnup capability of the steady-state code (FRAPCON) used for licensing audit calculations. (2) General capability (including high burnup) of the transient code (FRAPTRAN) used for special studies. (3) Adequacy at high burnup of fuel damage criteria used in regulation for reactivity accidents. (4) Adequacy at high burnup of models and fuel related criteria used in regulation for loss-of-coolant accidents (LOCAs). (5) Effect of high burnup on fuel system damage during normal operation, including control rod insertion problems. A distinction is made between technical issues, which may or may not have direct licensing impacts, and licensing issues. The RIAs became a licensing issue when the French test in CABRI showed that cladding failures could occur at fuel enthalpies much lower than a value currently used in licensing. Fuel assembly distortion became a licensing issue when control rod insertion was affected in some operating plants. In this presentation, these technical issues will be described and the NRC`s plan of action to address them will be discussed.

  2. OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

    International Nuclear Information System (INIS)

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  3. OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  4. TRIGA fuel burn-up calculations and its confirmation

    International Nuclear Information System (INIS)

    The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.

  5. 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.

  6. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  7. Performance of fast reactor irradiated fueled emitters at goal burnup

    International Nuclear Information System (INIS)

    UO2-fueled W emitters were examined that had been irradiated to goal burnups of approximately 4 at.% at emitter surface temperatures to 1820 K in a fast reactor to establish their performance for use in thermionic reactors with power levels from tens of kilowatts to multimegawatts. The examinations provided first-time data on structural integrity, dimensional stability, component compatibility, and fuel and fission product behavior. The data are consistent with similar measurements at approximately 2 at.% burnup with the exception of one emitter which breached the W during irradiation

  8. Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

    International Nuclear Information System (INIS)

    A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO2), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO2/UO2), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin

  9. Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

    International Nuclear Information System (INIS)

    For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B4C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability

  10. Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

    International Nuclear Information System (INIS)

    Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez-Lucuta model was favorable

  11. Need for higher fuel burnup at the Hatch Plant

    International Nuclear Information System (INIS)

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch's operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about

  12. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

    Beckhman, J.T. [Georgia Power Co., Birmingham, AL (United States)

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.

  13. Taking burnup credit for interim storage and transportation system for BWR fuels

    International Nuclear Information System (INIS)

    In the back-end issues of nuclear fuel cycle, selection of reprocessing or one-through is a big issue. For both of the cases, a reasonable interim storage and transportation system is required. This study proposes an advanced practical monitoring and evaluation system. The system features the followings: (l) Storage racks and transportation casks taking credit for burnup. (2) A burnup estimation system using a compact monitor with Cd- Te detectors and fission chambers. (3) A neutron emission-rate evaluation methodology, especially important for high burnup MOX fuels. (4) A nuclear materials management system for safeguards. Current storage system and transport casks are designed on the basis of a fresh fuel assumption. The assumption is too conservative. Taking burnup credit gives a reasonable design while keeping conservatism. In order to establish a reasonable burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of some modules such as TGBLA, ORIGEN, CITATION, MCNP and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. The code takes operational history such as, power density, void fraction into account. This code is applied to the back-end issues for a more accurate design of a storage and a transportation system. The ORIGEN code is well-known one-point isotope depletion code. In the calculation system, the code calculates isotope compositions using libraries generated from the TGBLA code. The CITATION code, the MCNP code, and the KENO code are three dimensional diffusion code, continuous energy Monte Carlo code, discrete energy Monte Carlo code, respectively. Those codes calculate k- effective of the storage and transportation systems using isotope compositions generated from the ORIGEN code. The CITATION code and the KENO code are usually used for practical designs. The MCNP code is used for reference

  14. Revaluation on measured burnup values of fuel assemblies by post-irradiation experiments at BWR plants

    International Nuclear Information System (INIS)

    Fuel composition data for 8x8 UO2, Tsuruga MOX and 9x9-A type UO2 fuel assemblies irradiated in BWR plants were measured. Burnup values for measured fuels based on Nd-148 method were revaluated. In this report, Nd-148 fission yield and energy per fission obtained by burnup analyses for measured fuels were applied and fuel composition data for the measured fuel assemblies were revised. Furthermore, the adequacies of revaluated burnup values were verified through the comparison with burnup values calculated by the burnup analyses for the measured fuel assemblies. (author)

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

    International Nuclear Information System (INIS)

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

  16. Burnup verification at Arkansas Nuclear One-unit 1 using the Fork measurement system

    International Nuclear Information System (INIS)

    The Fork measurement system, designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program, has been used for several years to examine spent fuel assemblies at nuclear reactors around the world. The objective of the test program described here is to demonstrate the ability of the Fork system to verify the records for assembly burnup at U.S. nuclear utilities. The measurements described here were performed at Arkansas Nuclear One, operated by Energy Operations, Inc. The Fork system was used to examine 34 assemblies in the storage pool of Arkansas Nuclear One-Unit 1. The correlation between the neutron measurements and the reactor records produced an average random deviation in burnup of 3.0% from the calibration, which translates into an average variation of 2.2% in the reactor records for burnup. The system proved to be compatible with utility operations

  17. Benchmark calculation with MOSRA-SRAC for burnup of a BWR fuel assembly

    International Nuclear Information System (INIS)

    The Japan Atomic Energy Agency has developed the Modular Reactor Analysis Code System MOSRA to improve the applicability of neutronic characteristics modeling. The cell calculation module MOSRA-SRAC is based on the collision probability method and is one of the core modules of the MOSRA system. To test the module on a real-world problem, it was combined with the benchmark program 'Burnup Credit Criticality Benchmark Phase IIIC.' In this program participants are requested to submit the neutronic characteristics of burnup calculations for a BWR fuel assembly containing fuel rods poisoned with gadolinium (Gd2O3), which is similar to the fuel assembly at TEPCO's Fukushima Daiichi Nuclear Power Station. Because of certain restrictions of the MOSRA-SRAC burnup calculations part of the geometry model was homogenized. In order to verify the validity of MOSRA-SRAC, including the effects of the homogenization, the calculated burnup dependent infinite multiplication factor and the nuclide compositions were compared with those obtained with the burnup calculation code MVP-BURN which had already been validated for many benchmark problems. As a result of the comparisons, the applicability of MOSRA-SRAC module for the BWR assembly has been verified. Furthermore, it can be shown that the effects of the homogenization are smaller than the effects due to the calculation method for both multiplication factor and compositions. (author)

  18. Non-destructive burnup determination of PWR spent fuel using Cs-134/Cs-137 and Eu-154/Cs-137

    International Nuclear Information System (INIS)

    Burnups for 36 points of five rods in the G23 assembly of Kori unit 1 have been determined on the basis of gamma-ray spectrometric measurement of two isotopic ratios, Cs-134/Cs-137 and Eu-154/Cs-137 in combination with the results calculated by the SCALE4.4 SAS2H module. Benchmarking of the SAS2H module has been done for the existing experimental data of Cs-13134, Cs-137 and Eu-154 isotopic compositions in PWR spent fuel. The gamma ray counts of two isotopic ratios have been corrected with their branching ratios, decay rates and energy dependent counting efficiencies in order to get true ratios. The energy dependent counting efficiencies have been determined as a quadratic equation based on the gamma ray counts for Cs-134 and Eu-154 at fourth energy points. Finally, burnups have been determined by putting true ratios of two isotopic ratios to their burnup-to-ratio fitting functions, respectively. Then the measured burnups have been compared with the declared burnup by the nuclear power plant. It is revealed that burnups determined from Cs-134/Cs-137 are agreeable with the declared burnups in most cases within about 12% error except a measuring point of C13, one of G23 fuel rods. In the case of Eu-154/Cs-137, the measured burnup is much lower than the declared burnup, which seems to be derived from system errors. (author)

  19. Evaluation and Selection of Boundary Isotopic Composition for Burnup Credit Criticality Safety Analysis of RBMK Spent Fuel Management

    International Nuclear Information System (INIS)

    The on-site wet-type spent fuel storage facility ISF-1 is currently used for interim storage of spent nuclear fuel removed from Chernobyl NPP power units. The results of ISF-1 preliminary criticality analyses demonstrated the need for using the burnup credit principle in nuclear safety analysis. This paper provides results from the selection and testing of computer codes for determining the isotopic composition of RBMK spent fuel. Assessment is carried out and conclusions are made on conservative approaches to fuel burnup credit in subsequent ISF-1 safety assessment. (author)

  20. Development of a Mobile CZT Detector System for Burnup Measurement of Spent Fuel Assembly and On-Site Application

    International Nuclear Information System (INIS)

    The advantages of mobile CdZnTe (CZT) detector for nuclear safeguard applications of spent fuel burnup inspection in assembly storage pond are compactness, low cost and ease of operations. In this work, a mobile detection system shield with tungsten alloy was designed and then performed on-site. Net count rate of the 662 keV line of 137Cs was produced linearly with burnup as experimental data simulations shows, in which the deviation from linearity is smaller than 9%. As a result, the feasibility of the method using CZT detector to monitor spent nuclear fuel assembly burnup in a fuel pond was validated. The results calculated with Monte Carlo procedure Geant4 can provide a theoretical guide for the further burnup measurement. (author)

  1. Experimental Comparison between High Purity Germanium and Scintillator Detectors for Determining Burnup, Cooling Time and Decay Heat of Used Nuclear Fuel

    OpenAIRE

    Jansson, Peter; Grape, Sophie; Tobin, Steve; Liljenfeldt, Henrik

    2014-01-01

    A experimental study of the gamma-ray energy spectra from used nuclear fuel has been performed. Four types of detectors were used to measure spectra from three PWR used fuel assemblies stored at the interim storage for used fuel in Sweden, CLAB: HPGe, LaBr3, NaI and BGO. The study was performed in the context of used fuel characterization for the back end of the fuel cycle in Sweden. Specifically, the purpose was to evaluate the behaviour of the different scintillator detectors (LaBr3, NaI an...

  2. Double-strata high burnup fuel performance in light water reactors

    International Nuclear Information System (INIS)

    This study is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. Two interrelated criteria, proliferation resistance and high-burnup, form the general framework of the fuel management scenario with the highest priority given to light water reactor technology and plutonium-free fresh fuel. Logically it implies the use of uranium oxide with enrichment close to 20%, whose effective utilization forms the main subject of the present paper. A sequence of two irradiation cycles for the same fuel pins in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140 GWd/tHM. Being as large as 8% in the final isotopic vector, the fraction of 238Pu serves as an inherent protective measure against plutonium proliferation

  3. Reevaluation of fuel enthalpy in NSRR test for high burnup fuels

    International Nuclear Information System (INIS)

    This paper describes the recent procedure of evaluation of the fuel enthalpy in the reactivity initiated accident (RIA) simulating tests performed at the nuclear safety research reactor (NSRR), and reports some important updates of the fuel enthalpies in the tests with high burnup PWR fuels. Previously, the fuel enthalpy had been evaluated by the procedure based on the short-life fission product measurement, i.e. a pellet slice was sampled from the test fuel rod after the NSRR test, a chemical separation process was applied to the solution of the pellet slice to separate barium, and the amount of Ba-140 was measured by gamma spectrometry on the separated barium. But a part of the results showed significant scattering even within the similar tests with similar fuels, which should have showed similar fuel enthalpies. The scattering appears to indicate the difficulty in treatment of the short-life nuclides after the completion of the NSRR test and unsuccessful measurement of the amount of fuel dissolved in the specimen preparation. Another difficulty of the procedure is that it is not repeatable for a specimen and so double check of an evaluation is not possible. Hence, an alternative procedure, which is based on the total amount of fissile materials evaluated by mass analysis, was developed and has been applied for the tests after 2003; the amount of fissile materials is input to a well-verified neutron transport calculation model for the NSRR reactor core to calculate a coupling factor of power densities between the test fuel rod and the NSRR driver fuel rods. This procedure does not require quickness and is repeatable, so it is applicable even many years later if the fuel sample is available. The recent procedure was thus applied to the tests before 2003, whose burnups are below 60 GWd/tU. It was shown that the fuel enthalpy had been significantly underestimated in the tests with high burnup PWR fuels: the test series HBO and TK. In this paper, the procedure

  4. Burnup credit methodology in the NPP Krzko spent fuel pool reracking project

    International Nuclear Information System (INIS)

    NPP Krzko is going to increase the capacity of the spent fuel storage pool by replacement of the existing racks with high-density racks. The design, rack manufacturing and installation has been awarded to the Framatome ANP GmbH. Burnup credit methodology, which has been already approved by the Slovenian Nuclear Safety Administration in previous licensing of existing racks, will be again implemented in the licensing process with the recent methodology improvements. Specific steps of the criticality analysis and representative results are presented in the paper showing also the current national practice of the burnup credit implementation. (author)

  5. Studies at INR-Pitesti for developing fuels of high burnup suitable to CANDU 6 reactor

    International Nuclear Information System (INIS)

    Increasing burnup allows the utility to get the same kWh output with a diminished tonnage of fissile material and provides a saving in the cost of fuel manufacturing as well as of spent fuel disposal. The RU, SEU, MOX, DUPIC fuel cycles and CANFLEX fuel bundles concept compatible with CANDU 6 reactor are presented. INR projects for developing SEU 43 fuel bundles supported by IAEA-Vienna are also presented. Particularly, one gives an overlook of standard CANDU and advanced SEU 43 nuclear fuel cycles. The paper presents also the current and future directions of studies implied by the research program in the nuclear fuel field of RAAN (The Autonomous Authority for Nuclear Activities). Among these, mentioned are: working out of the manual of physics of CANDU core with slightly enriched uranium; technological studies aiming at reducing the effects of limiting factors of the fuel lifetime and at burnup extension; obtaining new fuels as vectors of advanced cycles; off reactor tests of SEU 43 clusters; in-reactor tests of SEU 43 experimental fuel elements; developing computer codes for analysis of SEU, MOX and DUPIC fuel behavior; in-reactor tests of experimental MOX and DUPIC elements

  6. Application of burnup credit with partial boron credit to PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    The outcome of performing a burnup credit criticality safety analysis of a PWR spent fuel storage pool is the determination of burnup credit loading curves BLC=BLC(e) for the spent fuel storage racks designed for burnup credit, cp. Reference. A burnup credit loading curve BLC=BLC(e) specifies the loading criterion by indicating the minimum burnup BLC(e) necessary for the fuel assembly with a specific initial enrichment e to be placed in storage racks designed for burnup credit. (orig.)

  7. Research on Integrity of High Burnup Spent Fuel Under the Long Term Dry Storage

    International Nuclear Information System (INIS)

    Objectives were to acquire the following behaviour data by dynamic load impact tests on high burnup spent fuel rods of BWR and PWR and to improve the guidance of regulation of spent fuel storage and transportation. (1) The limit of load and strain for high burnup fuel in the cask drop accident. (2) The amount of deformation of high burnup fuel rods under dynamic load impact. (3) The amount of fuel pellet material released from fuel rods under dynamic load impact

  8. Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

    International Nuclear Information System (INIS)

    The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.

  9. Methods of RECORD, an LWR fuel assembly burnup code

    International Nuclear Information System (INIS)

    The RECORD computer code is a detailed rector physics code for performing efficient LWR fuel assembly calculations, taking into account most of the features found in BWR and PWR fuel designs. The code calculates neutron spectrum, reaction rates and reactivity as a function of fuel burnup, and it generates the few-group data required for use in full scale core simulation and fuel management calculations. The report describes the methods of the RECORD computer code and the basis for fundamental models selected, and gives a review of code qualifications against measured data. (Auth. /RF)

  10. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  11. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    International Nuclear Information System (INIS)

    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 UO2 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.

  12. 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.

  13. Fuel burnup extension effect on the fuel utilization and economical impact for a typical PWR plant

    International Nuclear Information System (INIS)

    Currently in Japan, fuel assembly average burn-up is limited to 48GWd/t and is going to be extended to 55GWd/t in these years. Moreover, R and D programs for further extension are under operation. Simultaneous extension of fuel burn-up limitation and cycle length reduces the number of fuel required to produce a given amount of energy reducing the radioactive waste generation, the occupational radiation exposure and the electricity generation cost. In this paper, the effect of fuel burn-up and operation cycle length extension is estimated from the view point of electricity generation cost and amount of discharged fuel assemblies, and the desirable burn-up extension in the future is studied. The present 5wt% uranium-235 enrichment restriction for commercial reactors divides the burn-up extension implementation in two steps. The fuel burn-up achievable with the present 5wt% enrichment limitation and without it is analyzed. A standard 3 loop PWR plant loading 17x17 fuel assemblies has been chosen for the feasibility study of operation cycle longer than 15 months and up to 24 months under extended fuel burn-up limitation. With the 5wt% enrichment limitation, the maximum assembly average burn-up is between 60GWd/t and 70GWd/t. Three batches reload fuel strategy and 18 months operation cycle allow the electricity generation cost reduction in about 4% and the number of fuel assemblies discharged per year is reduced in approximately 15% compared with the current 48GWd/t fuel. Relaxing the enrichment limitation, for the 24 months operation cycle with 3 batches reload fuel strategy, the maximum assembly average burn-up become 80GWd/t. The electricity generation cost reduction is about 8% and the number of fuel assemblies discharged per year is reduced in approximately 35% compared with the current condition. This study shows the contribution of simultaneous extension of fuel burn-up limitation and operation cycle length to reduce the electricity generation cost and the number

  14. 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.

  15. A relative risk comparison of criticality control strategies based on fresh fuel and burnup credit design bases

    International Nuclear Information System (INIS)

    The fresh fuel design basis provides some margin of safety, i.e., criticality safety is almost independent of loading operations if fuel designs do not change significantly over the next 40 years. However, the design basis enrichment for future nuclear fuel will most likely vary with time. As a result, it cannot be guaranteed that the perceived passivity of the concept will be maintained over the life cycle of a future cask system. Several options are available to ensure that the reliability of a burnup credit system is comparable to or greater than that of a system based on a fresh fuel assumption. Criticality safety and control reliability could increase with burnup credit implementation. The safety of a burnup credit system could be comparable to that for a system based on the fresh fuel assumption. A burnup credit philosophy could be implemented without any cost-benefit tradeoff. A burnup credit design basis could result in a significant reduction in total system risk as well as economic benefits. These reductions occur primarily as a result of increased cask capacities and, thus, fewer shipments. Fewer shipments also result in fewer operations over the useful life of a cask, and opportunities for error decrease. The system concept can be designed such that only benefits occur. These benefits could include enhanced criticality safety and the overall reliability of cask operations, as well as system risk and economic benefits. Thus, burnup credit should be available as an alternative for the criticality design of spent fuel shipping casks

  16. Reactivity loss validation of high-burnup PWR fuels with pile-oscillation experiments in MINERVE

    International Nuclear Information System (INIS)

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

  17. High Burnup UO2 Fuel Pellets with Dopants for WWER

    International Nuclear Information System (INIS)

    The currently achieved level of design and technology developments provided for the implementation of the fuel cycle (4x1) in WWER at the maximal design burnup of 56 MW.day/kgU per FA. Presently in Russia the program is under way to improve the technical and economic parameters of WWER fuel cycles characterized by an increased fuel usability. To meet the requirements placed on the new fuel that ensures the reliable operation under conditions of higher burnups complex activities are under way to optimize the composition and microstructure of fuel pellets as applied to WWER. This paper describes a general approach to providing the stimulated composition and microstructure of fuel via introducing various dopants. Aside from this, the paper presents the experimentally results of studies into the main technologic and operational characteristics of dopant containing fuel pellets including higher grain sizes, pores distribution and oxygen to metal ratio. The results of the experiments made it possible to work out the pilot commercial process of the modified fuel fabrication, to manufacture pellet batches to be semi-commercially operated at NPP with WWER. (author)

  18. Transient behaviour of high burnup fuel. Status report

    International Nuclear Information System (INIS)

    This Status Report is a follow-on to the CSNI Specialist Meeting on Transient Behaviour of High Burnup Fuel which was held in Cadarache, France, from September 12. to 14., 1995. The Status Report identifies the needs and rationale for any further work to better understand the transient behaviour of high burnup fuel. The different options to perform that work, from analytical to experimental activities, and discussion on the potential benefits of performing new integral tests are also addressed. A brief description of the major on-going and short-term planned activities in this field is included as additional information. The main conclusions from this effort are highlighted. (K.A.)

  19. Destructive radiochemical analysis of uraniumsilicide fuel for burnup determination

    Energy Technology Data Exchange (ETDEWEB)

    Gysemans, M.; Bocxstaele, M. van; Bree, P. van; Vandevelde, L.; Koonen, E.; Sannen, L. [SCK-CEN, Boeretang, Mol (Belgium); Guigon, B. [CEA, Centre de Cadarache, Saint Paul lez Durance (France)

    2004-07-01

    During the design phase of the French research reactor Jules Horowitz (RJH) several types of low enriched uranium fuels (LEU), i.e. <20% {sup 235}U enrichment, are studied as possible candidate fuel elements for the reactor core. One of the LEU fuels that is taken into consideration is an uraniumsilicide based fuel with U{sub 3}Si{sub 2} dispersed in an aluminium matrix. The development and evaluation of such a new fuel for a research reactor requires an extensive testing and qualification program, which includes destructive radiochemical analysis to determine the burnup of irradiated fuel with a high accuracy. In radiochemistry burnup is expressed as atom percent burnup and is a measure for the number of fissions that have occurred per initial 100 heavy element atoms (%FIMA). It is determined by measuring the number of heavy element atoms in the fuel and the number of atoms of selected key fission products that are proportional to the number of fissions that occurred during irradiation. From the few fission products that are suitable as fission product monitor, the stable Nd-isotopes {sup 143}Nd, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148Nd}, {sup 150}Nd and the gamma-emitters {sup 137}Cs and {sup 144}Ce are selected for analysis. Samples form two curved U{sub 3}Si{sub 2} plates, with a fuel core density of 5.1 and 6.1 g U/cm{sup 3} (35% {sup 235}U) and being irradiated in the BR2 reactor of SCK x CEN{sup [1]}, were analyzed. (orig.)

  20. Application of depletion perturbation theory to fuel cycle burnup analysis

    International Nuclear Information System (INIS)

    Over the past several years static perturbation theory methods have been increasingly used for reactor analysis in lieu of more detailed and costly direct computations. Recently, perturbation methods incorporating time dependence have also received attention, and several authors have demonstrated their applicability to fuel burnup analysis. The objective of the work described here is to demonstrate that a time-dependent perturbation method can be easily and accurately applied to realistic depletion problems

  1. The implementation of burnup credit in VVER-440 spent fuel

    International Nuclear Information System (INIS)

    The countries using Russian reactors VVER-440 cooperate in reactor physics in Atomic Energy Research (AER). One of topic areas is 'Physical Problems of Spent Fuel, Radwaste and Decommissioning' (Working Group E). In this article, in the first part is an overview about our activity for numerical and experimental verification of codes which participants use for calculation of criticality, isotopic concentration, activity, neutron and gamma sources and shielding is shown. The set of numerical benchmarks (CB1, CB2, CB3 and CB4) is very similar (the same idea, the VVER-440) to the OECD/NEA/NSC Burnup Credit Criticality Benchmarks, Phases 1 and 2. In the second part, verification of the SCALE 4.4 system (only criticality and nuclide concentrations) for VVER-440 fuel is shown. In the third part, dependence of criticality on burnup (only actinides and actinides + fission products) for transport cask C30 with VVER-440 fuel by optimal moderation is shown. In the last part, current status in implementation burnup credit in Slovakia is shown. (author)

  2. Use of burnup credit in criticality safety design analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    temperature and density, presence of soluble boron in the core (PWR), use of fixed neutron absorbers (control rods, burnable poison rods, axial power shaping rods), use of integral burnable absorbers (gadolinium or erbium bearing fuel rods, IFBA rods). It will be shown how a bounding approach can be obtained for the impact of these parameters on the reactivity of the storage system. The criticality calculation procedure consists in the following main steps: Isotopic selection and validation; Validation of the criticality calculation code applied; Sensitivity studies on the reactivity effects of axial and horizontal burnup profiles of fuel assemblies; Determination of the criticality acceptance criterion (maximum allowable neutron multiplication factor including the impacts of all the mechanical and calculational uncertainties) and determination of the loading curve. The fundamentals of isotopic selection will be defined, and a survey of the benchmark experiments available for isotopic validation and validation of the criticality calculation code applied will be given. Since the parameters and conditions characterizing the benchmark experiments are usually different from the parameters and conditions describing the spent fuel storage system of interest, a method of checking the applicability of such experiments to the storage system will be briefly described. This method bases the applicability on the similarity of sensitivity coefficients which are defined for the underlying nuclear data characterizing the isotopic compositions and their effect on the spent fuel reactivity. The fact that the axial burnup distribution in a fuel assembly is non-uniform must be considered in the analysis of the storage system. The difference between the system's neutron multiplication factor obtained by using an axially varying burnup profile and the system's neutron multiplication factor obtained by assuming a uniform distribution of the averaged burnup of this profile is known as the 'end

  3. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  4. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  5. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  6. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    International Nuclear Information System (INIS)

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup

  7. Surface harmonics method for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel

    International Nuclear Information System (INIS)

    Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)

  8. Surface harmonics method for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V. F.; Davidenko, V. D.; Polismakov, A. A.; Tsibulsky, V. F. [Russian Research Center Kurchatov Inst., Nuclear Reactor Inst., 123182, Moscow (Russian Federation)

    2006-07-01

    Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)

  9. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  10. Study on burnup credit evaluation method at JAERI towards securing criticality safety rationale for management of spent fuel

    International Nuclear Information System (INIS)

    A higher initial 235U enrichment is currently required in the nuclear fuel fabrication specification to realize higher fuel burnup. Traditionally, in the criticality safety design of spent fuel (SF) storage and transportation (S/T) casks or facilities, the fuel is usually assumed to be at its full initial enrichment (so called fresh fuel assumption) to provide a large safety allowance, which is sometimes excessively given, for example requiring unnecessarily large space between fuel assemblies. The burnup credit taken for criticality safety design is firstly implemented to the SF Storage Rack of Rokkasho Reprocessing Facility, which is completed and expected for operation soon. Except for that, no burnup credit has been taken in criticality safety design for SF S/T casks or intermediate storage facilities in Japan. Since in the near future it is considered inevitable to handle spent fuel massively, it is desired to implement the rational S/T design saving safety and economy by taking into account the fuel burnup in the criticality safety control. Computer codes and data which are vital to assess criticality safety in the design stage of nuclear fuel cycle facility have been developed and prepared to constitute a Japanese criticality safety handbook at JAERI

  11. In-core fuel management amd attainable fuel burn-up in TRIGA

    International Nuclear Information System (INIS)

    The principles of in-core fuel management in research reactors, and especially in TRIGA, are discussed. Calculations made to determine the attainable fuel burn-up values of various fuel element types in the Otaniemi TRIGA Mark II reactor are described and the results obtained are given. Recommendations are given of how to perform the in-core fuel management to achieve good fuel utilization. The results obtained indicate that burn-up values of up to 5 and 2.5 MWd/element can be achieved for the 8 wt-% U Al clad and the 8.5 wt-% U SS clad elements, respectively. (author)

  12. 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.

  13. Chemical form of fission products in high burnup fuels

    International Nuclear Information System (INIS)

    In order to make a proper assessment of candidate materials for advanced high-burnup fuels, thermochemical studies of fuel materials have been performed. Using data from the ECN thermochemical database (TBASE), which has been updated and extended for the present work, the suitability of various advanced fuel materials and inert matrices is studied. Detailed thermodynamic equilibrium calculations are performed for Pu0.42U0.58O2 and Pu0.40U0.60N for values of the burnup up to 200 MWd/kgHM. The formation of metallic phases, the pressure buildup and the stability of nitride or oxide phases is studied for each fuel type. The results for the chemical form of the solid fission products are given. The chemical aspects of the use of the inert matrix spinel (MgAl2O4) in combination with oxide fuel will be discussed. Experimental research on the compatibility of various types of inert matrices (nitrides, spinel) is in progress at ECN. (author)

  14. Effects of axial burnup distributions on the reactivity of spent fuel

    International Nuclear Information System (INIS)

    Criticality safety analyses for spent fuel shipping casks will eventually need to take credit for the decreased reactivity of spent fuel assemblies resulting from burnup. In order to do so, it will be necessary to assess the reactivity effects of the multitude of burnup shapes that can characterize spent fuel. A computer program, CASAX, has been written that allows the analyst to quickly evaluate the reactivity effects of actual and simplified axial burnup distributions on a group of PWR fuel assemblies. CASAX employs one dimensional, two group diffusion calculations to determine the k-effective of a cluster of assemblies. Assembly average, burnup dependent, two group cross sections for CASAX were obtained from CASMO3 using physical properties representative of Westinghouse 17 x 17 assemblies. Reactivity results are presented in terms of (k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup)/(k for a uniform axial burnup distribution at the assembly average burnup). Axial burnup distributions can have both positive and negative effects on the calculated k-effective. Positive reactivity effects generally result at high assembly average burnups and for axial distributions with low burnups in the assembly's tips

  15. Effect of burnup dependence of fuel cladding gap properties on WWER core characteristics

    International Nuclear Information System (INIS)

    Dependence of gas gap properties on burnup has been obtained with use of TRANSURANUS code. Implemented dependency on burnup is based on TRANSURANUS calculations of different fuel pins upon different linear power Ql. Obtained dependence was implemented into DYN3D code and results of new dependence effect on characteristics of WWER fuel loadings are presented. The work was performed in framework of orders BMU SR 2511 and BMU R0801504 (SR2611). The report describes the opinion and view of the contractor-State Scientific and Technical Centre on Nuclear and Radiation Safety-and does not necessarily represent the opinion of the ordering party-BMU-BfS/GRS and TUEV SUED. (Authors)

  16. Framatome-ANP extended burnup experience and views on LWR fuels

    International Nuclear Information System (INIS)

    In every sense of the term, nuclear fuel forms the core of nuclear power plants (NPPs). Although there are many equipment items important for their safety function or for their participation in NPP availability, the fuel, in essence renewable, is one of the key elements which have to be acted upon if utilities are to be helped to fulfil their mission of generating power in total safety and supplying the kWh to their customers at the best price. Nuclear fuel is also the core business of the Framatome-ANP Fuel Business Group: pooling and rationalising the available skills - technical, cultural and human - supplied by each of the partners forms a challenge which it is up to each and every one to meet in a cooperative spirit. This paper gives an outline of the company's extended burnup experience, current R and D, and its plans for the future. (author)

  17. The burnup dependence of light water reactor spent fuel oxidation

    International Nuclear Information System (INIS)

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO2 to higher oxides. The oxidation of UO2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO2 to UO2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

  18. Effect of fuel burnup history on neutronic characteristics of WWER-1000 core

    International Nuclear Information System (INIS)

    The paper analyzes fuel burnup history effect on neutronic characteristics of WWER-1000 core with use of the DYN3D codes. The DYN3D code employs the local Pu-239 concentration as an indicator of burnup spectral history. The calculations have been performed for the first four fuel loadings of Khmelnitsky NPP unit 2 and stationary fuel loading with TVSA. The effect of fuel burnup history is shown both on macro-characteristics on the reactor core and on local values of burnup and power

  19. Microstructural characterization of high burn-up mixed oxide fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Teague, Melissa, E-mail: melissa.teague@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Gorman, Brian; King, Jeffrey [Colorado School of Mines, 1500 Illinois St, Golden, CO 80401 (United States); Porter, Douglas; Hayes, Steven [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2013-10-15

    High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

  20. MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis

    International Nuclear Information System (INIS)

    The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2, and the Weapons-Grade Mixed Oxide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data

  1. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  2. Nondestructive analysis of RA reactor fuel burnup, Program for burnup calculation base on relative yield of 106Ru, 134Cs and 137Cs in the irradiated fuel

    International Nuclear Information System (INIS)

    Burnup of low enriched metal uranium fuel of the RA reactor is described by two chain reactions. Energy balance and material changes in the fuel are described by systems of differential equations. Numerical integration of these equations is base on the the reactor operation data. Neutron flux and percent of Uranium-235 or more frequently yield of epithermal neutrons in the neutron flux, is determined by iteration from the measured contents of 106Ru, 134Cs and 137Cs in the irradiated fuel. The computer program was written in FORTRAN-IV. Burnup is calculated by using the measured activities of fission products. Burnup results are absolute values

  3. Dissolution studies of spent nuclear fuels

    International Nuclear Information System (INIS)

    To obtain quantitative data on the dissolution of high burnup spent nuclear fuel, dissolution study have been carried out at the Department of Chemistry, JAERI, from 1984 under the contract with STA entitled 'Reprocessing Test Study of High Burnup Fuel'. In this study PWR spent fuels of 8,400 to 36,100 MWd/t in averaged burnup were dissolved and the chemical composition and distribution of radioactive nuclides were measured for insoluble residue, cladding material (hull), off-gas and dissolved solution. With these analyses basic data concerning the dissolution and clarification process in the reprocessing plant were accumulated. (author)

  4. 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

  5. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Russian Research Centre, Moscow (Russian Federation)

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  6. Investigation of several methods to set burnup for criticality safety assessment of spent fuel transport casks

    International Nuclear Information System (INIS)

    Several currently available methods to set burnup for depletion calculation are reviewed and discussed about its adequacy for criticality safety assessment of spent fuel (SF) transport casks by taking burnup credit (BC) into accounts. Various errors associated with BC criticality analyses are evaluated and converted to equivalent burnup to compare each other. Methods are proposed to use some reduced burnups equivalent to compensation of these associated errors. Effects of assumption of axial burnup distribution on criticality calculation and irradiation history parameter variation on depletion calculation are evaluated with OECD/NEA BC international benchmark data. (author)

  7. Fuel Element Designs for Achieving High Burnups in 220 MW(e) Indian PHWRs

    International Nuclear Information System (INIS)

    Presently 19-element natural uranium fuel bundles are used in 220 MW(e) Indian PHWRs. The core average design discharge burnup for these bundles is 7000 MW·d/Te U and maximum burnup for assembly goes upto of 15 000 MWD/Te U. Use of fuel materials like MOX, Thorium, slightly enriched uranium etc in place of natural uranium in 19-element fuel bundles, in 220 MW(e) PHWRs is being investigated to achieve higher burnups. The maximum burnup investigated with these bundles is 30 000 MW·d/Te U. In PHWR fuel elements no plenum space is available and the cladding is of collapsible type. Studies have been carried out for different fuel element target burnups with different alternative concepts. Modification in pellet shape and pellet parameters are considered. These studies for the PHWR fuel elements/assemblies have been elaborated in this paper. (author)

  8. Burn-up measurements at TRIGA fuel elements containing strong burnable poison

    International Nuclear Information System (INIS)

    The reactivity method of determining the burn-up of research reactor fuel elements is applied to the highly enriched FLIP elements of TRIGA reactors. In contrast to other TRIGA fuel element types, the reactivity of FLIP elements increases with burn-up due to consumption of burnable poison. 33 fuel elements with burn-up values between 3% and 14% were investigated. The experiments showed that variations in the initial fuel composition significantly influence the reactivity and, consequently, increase the inaccuracy of the burn-up measurements. Particularly important are variations in the initial concentration of erbium, which is used as burnable poison in FLIP fuel. A method for reducing the effects of the material composition variations on the measured reactivity is presented. If it is applied, the accuracy of the reactivity method for highly poisoned fuel elements becomes comparable to the accuracy of other methods for burn-up determination. (orig.)

  9. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  10. Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1993-06-01

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author).

  11. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The purpose of this meeting was to review the ''state-of-the-art'' in the area of Fuel Performance at High Burnup for Water Reactors. Previous IAEA meetings on this topic were held in Mol in 1981 and 1984 and on related topics in Stockholm and Lyon in 1987. Fifty-five participants from 16 countries and two international organizations attended the meeting and 28 papers were presented and discussed. The papers were presented in five sub-sessions and during the meeting, working groups composed of the session chairmen and paper authors prepared the summary of each session with conclusions and recommendations for future work. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  12. Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

    International Nuclear Information System (INIS)

    Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a 'best estimate' value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 fission products (FPs) of PWR-MOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF- 3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections. (authors)

  13. Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

    Directory of Open Access Journals (Sweden)

    Lecarpentier D.

    2013-03-01

    Full Text Available Burnup Credit (BUC is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a “best estimate” value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library. Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.

  14. Burnup behavior of FBR fuels sourced in uranium and plutonium recycled in PWRs and its influence on fuel cycle economy

    International Nuclear Information System (INIS)

    to established the same burnup performance, mainly because of poisonous influence of 236U. The problem of 236U is, rather than its neutron absorptivity, its transformation to 237Np having a large cross-section area of neutron capture. The MOX fuels manufactured with the matrixes sourced in the recovery uranium, however, can compensate the neutron toxicity of 236U with 235U enriched in the matrixes while contributing to core burnup. So, these MOX fuels require a decrease of plutonium, compared with the MOX fuel made with the depleted natural uranium. The quantitative fuel reproduction in the first recycle was estimated as follows: In the case of A-type matrix, the MOX fuels for 13.7 PWRs and the reformed uranium fuels for 12.4 PWRs can be obtained from the conventional spent fuels of 100 PWRs. In the case of B-type matrix, the MOX fuels for 13.2 PWRs and the reformed fuels for 14.2 PWRs can be manufactured from the same spent fuels. In the case of C-type matrix, the MOX fuels and the reformed uranium fuels can be provided for 12.7 and 14.2 PWRs, respectively. Therefore, the fuel recycle efficiencies in the PWR system are 26.1%, 27.4 % and 26.9 % for the cases A, B and C, respectively. (partially reported in 4th RRTD Int. Workshop for Asian Nucl. Prospect (1st Asian Nuclear Prospect Workshop), October 19-21, 2008 Kobe, Japan.) The multi-recycle in the light water reactor (LWR) fuel system, however, will be involved by that 236U in reformed fuels increases more and more with recycle times. The spent-fuel resources are recommended to be utilized in the fast breeder reactor (FBR) fuel cycle. Thus, we have been studying the effective use of uranium and plutonium resources in the FBR cycle cooperating with the LWR system. In this work, a FBR of practical type (1.5 GWe-class) was conceptually designed in the SRAC numerical system, referring to a commercial-type FBR model proposed by Japan Atomic Energy Agency and The Japan Atomic Power Co. Its main design parameters are

  15. Dissolution of low burnup Fast Flux Test reactor fuel

    International Nuclear Information System (INIS)

    The first Fast-Flux Test Facility reactor fuel [mixed (U,Pu)O2 composition] has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 19970C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 950C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 290C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 950C dissolution contained the equivalent of 198 mg of 239Pu per 100 g of hulls, while the cladding from the 290c experiments contained only 0.21 mg of 239Pu per 100 g of hulls. 9 references, 5 figures

  16. Nuclear fuel

    International Nuclear Information System (INIS)

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts. (Kako, I.)

  17. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  18. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  19. High fuel burn-up and nonproliferation in PWR-type reactor on the basis of modified Th-fuel

    International Nuclear Information System (INIS)

    Neutronics-physical characteristics of the fuel lattice of a PWR-type reactor cooled by light water and by a mixture of light and heavy water have been analyzed. Th-fuel containing an essential amount of 231Pa and 232U is used, which allows an increase in fuel burn-up by a factor of 2-5 compared with that of traditional oxide uranium fuel with light water. It is important to underline that this is attained under the negative coolant density reactivity effect using cross sections of 231Pa and 232U from the updated JENDL-3.2 nuclear library. This radical increase of fuel burn-up is accompanied by a small change of reactivity during fuel irradiation (K∞=1.1 / 1.0), that favorably affects safety parameters of the reactor operation. A considerable percentage of 232U in fuel, and consequently in U, is a strong barrier against the proliferation of such weapon nuclide as 233U. (authors)

  20. WWER-1000 fuel cycle economical improvement by reaching high fuel burnup

    International Nuclear Information System (INIS)

    The use of some Fuel Assemblies (FAs) of conventional design in 4 fuel cycles has allowed to increase unloaded FA average burnup from 38.5-40 MWd/kgU up to 42-43 MWd/kgU. It makes it possible to reduce spent fuel amount and, respectively, the fuel cycle back end costs. The increase of Ukrainian WWER-1000 fuel burnup has not decreased the fuel reliability. The number of leaking, unloaded ahead of schedule FAs, FAs having reached the failure criterion, differ substantially (several times). Correspondingly the share of leaking FAs (FPLR) and the share of failed FAs (FPFR) will be different. Average value of FPFR calculated according to the number of unloaded ahead of schedule FAs for Ukrainian WWER-1000 (except Rovno-3) not more than (0.6-1)x10-5 (6-10 ppm). Average fuel pin leaking rate for Ukrainian WWER-1000 (without taking into account some cases, Rovno-3) corresponds to 2 x10-5 -2.8 x10-5 (20-28) ppm. The experience of fuel operation and principal results of the irradiated FAs examination allow to accept a possibility of further fuel burnup increase. The situation regarding fuel reliability operation on Rovno-3 units requires further analysis and additional measures

  1. Nuclear fuel

    International Nuclear Information System (INIS)

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.)

  2. Basic safety research for high burnup fuels in light water reactors

    International Nuclear Information System (INIS)

    While the high burnup programs for LWR'S fuel have been proceeding, the piling up fundamental data on high burnup fuels is more desirable to conduct even more precise assessment of fuel integrity and safety. This study covers the influence of gadelinia addition on fuel pellets, of hydrogen absorption on zircaloy waterside corrosion, of fission products on pellet-cladding interaction and reactor structual materials. (author)

  3. Burn-up credit criticality benchmark. Phase 4-B: results and analysis of MOX fuel depletion calculations

    International Nuclear Information System (INIS)

    The DECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide (UOX) fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The exercises consisted of inventory calculations of MOX fuels for two initial plutonium compositions. The depletion calculations were carried out using three representations of the MOX assemblies and their interface with UOX assemblies. This enabled the investigation of the spatial and spectral effects during the irradiation of the MOX fuels. (author)

  4. Implementation of burnup credit in spent fuel management systems. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report

  5. Pool inspection techniques for surveillance and further development of high burnup fuel assemblies

    International Nuclear Information System (INIS)

    The pool inspection techniques which have been used in fuel assembly surveillance programs for many years are suitable for high burnup fuel assemblies too. The techniques have been adapted to the requirements of new fuel assembly concepts with higher burnup potential. For high burnup, the emphasis within the scope of examination techniques available has shifted towards a characterization of the corrosion behaviour and surveillance of the geometrical dimensions of the fuel assemblies. In order to accomplish these tasks complementary techniques will have to be developed. (orig.)

  6. An analysis of burnup reactivity credit for reactor RA spent fuel storage

    International Nuclear Information System (INIS)

    The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

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

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

    International Nuclear Information System (INIS)

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

  9. High burn-up structure of U(Mo) dispersion fuel

    Science.gov (United States)

    Leenaers, A.; Van Renterghem, W.; Van den Berghe, S.

    2016-08-01

    The evolution of the high burn-up structure (HBS) in U(Mo) fuel irradiated up to a burn-up of ∼70% 235U or ∼5 × 1021 f/cm3 or ∼120 GWd/tHM is described and compared to the observation made on LWR fuel. Scanning and transmission electron microscopy was performed on several samples having different burn-ups in order to get a better understanding of the mechanisms leading to the high burn-up structure formation. Even though there are some substantial differences between the irradiation of ceramic and U(Mo) alloy fuels (crystal structure, enrichment, irradiation temperature …), it was found that in both fuels recrystallization initiates at the same threshold and progresses in a similar way with increasing fission density. In case of U(Mo), recrystallization leads to accelerated swelling of the fuel which could result in instability of the fuel plate.

  10. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

    International Nuclear Information System (INIS)

    This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan

  11. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C V; DeHart, M D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2000-03-13

    This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan.

  12. Variants of closing the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Andrianova, E. A., E-mail: Andrianova-EA@nrcki.ru; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  13. Variants of closing the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed

  14. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

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

    International Nuclear Information System (INIS)

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

  16. Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately

  17. Fission gas release and fuel rod chemistry related to extended burnup

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review the state of the art in fission gas release and fuel rod chemistry related to extended burnup. The meeting was held in a time when national and international programmes on water reactor fuel irradiated in experimental reactors were still ongoing or had reached their conclusion, and when lead test assemblies had reached high burnup in power reactors and been examined. At the same time, several out-of-pile experiments on high burnup fuel or with simulated fuel were being carried out. As a result, significant progress has been registered since the last meeting, particularly in the evaluation of fuel temperature, the degradation of the global thermal conductivity with burnup and in the understanding of the impact on fission gas release. Fifty five participants from 16 countries and one international organization attended the meeting. 28 papers were presented. A separate abstract was prepared for each of the papers. Refs, figs, tabs and photos

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

    International Nuclear Information System (INIS)

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

  19. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    International Nuclear Information System (INIS)

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels

  20. End effect analysis with various axial burnup distributions in high density spent fuel storage racks

    International Nuclear Information System (INIS)

    Highlights: • Criticality tests are carried out with various axial burnup distributions of fuel assemblies for spent fuel storage racks. • KENO-Va code system was used to obtain criticalities with 10 axial segments. • ORIGEN-S code system was used to obtain burnup dependent axial compositions. • The criticality and burnup dependent reactivity difference are obtained from the results. • End effect quantifications are satisfactory confirming the previous suggestions. - Abstract: End effect of spent fuel comes from the difference between uniform and actual axial burnup distributions of fuel assemblies. It is significant to control the criticality safety in spent fuel storage and transportation. This work is focused on estimation of end effect in the spent fuel of light water reactor for the spent fuel storage rack region-II. High and low burnups of corresponding different uranium enrichments are taken into consideration to analyze the end effect with different axial burnup distributions such as uniform, MOC and EOC profiles. Two types of fuel assemblies such as CE type and Westinghouse type are considered. The whole calculations have been carried out by using the SCALE6 code including ORIGEN-S and KENO-Va

  1. Burn-up credit criticality benchmark. Phase 4-A: reactivity prediction calculations for infinite arrays of PWR MOX fuel pin cells

    International Nuclear Information System (INIS)

    The OECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The report summarises and analyses the solutions to the specified exercises provided by 37 contributors from 10 countries. The exercises were based upon the calculation of infinite PWR fuel pin cell reactivity for fresh and irradiated MOX fuels with various MOX compositions, burn-ups and cooling times. In addition, several representations of the MOX fuel assembly were tested in order to check various levels of approximations commonly used in reactor physics calculations. (authors)

  2. Passive neutron assay of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it is found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, it is found that the neutron emission rate is proportional to burnup raised to 3.4 power. Recent investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies. 10 refs

  3. Passive neutron assay of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it was found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, the neutron emission rate is proportional to burnup raised to 3.4 power. The investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies

  4. Investigation of fuel rod behaviour under extended 1 burnup conditions with ROFEM fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Moscalu, D.R.; Popescu, I.A

    1998-06-01

    Extending burnup is a practical way to improve the economics of water-reactor operation, via enhanced fuel utilisation and reduced spent fuel volume. A dedicated fuel behaviour modeling computer code (entitled ROFEM-1B) has been developed in order to analyse high burnup fuel performance. The code was benchmarked on an experimental data base which include a significant number of irradiation experiments performed in TRIGA-INR Pitesti research reactor. Five fuel rod behaviour during irradiation up to 50 MWd kg{sup -1}UO{sub 2}{sup -1} burnup have been analysed by the code in the framework of the first phase of the international FUMEX code exercise co-ordinated by IAEA Vienna. The input experimental data package has been prepared by IFE-OECD Halden, Norway laboratory. In the second phase of the FUMEX exercise the participants have analysed eight simplified theoretical cases. The paper presents and discuss the results obtained with ROFEM-1B in this exercise and the comparison between code predictions and experimental data.

  5. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  6. Separation of Molybdenum From Spent Fuel Solution in Burnup Measurements Process

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    In order to establish a kind of automatic radiochemistry separation procedure of nuclide 100Mo from spent fuel solution in burnup measurements process, a method of separating Mo quickly and effectively from the feed solution is needed. In the studies,

  7. Supercell burnup model for the physics design of BWR fuel assemblies

    International Nuclear Information System (INIS)

    A code called SUPERB has been developed for the BWR fuel assembly burnup analyses using supercell model. Each of the characteristic heterogeneities of a BWR fuel assembly like water gap, poisoned pins, control blade etc., is treated by invoking appropriate supercell concept. The burnup model of SUPERB is so devised as to strike a balance between accuracy and speed. This is achieved by building isotopic densities in each fuel pin separately while the depletion equations are solved only in a few groups of pins or burnup zones and the multigroup neutron spectra are differentiated in fewer group of pincell types. Multiple fuel ring burnup is considered only for Gd isotopes. A special empirical formula allows the microscopic cross section of Gd isotopes to be varied even during burnup integration. The supercell model has been tested against Monte Carlo results for the fresh cold clean Tarapur fuel assembly with two Gd fuel pins. The burnup model of SUPERB has been validated against one of the most sophisticated codes LWR-WIMS for a benchmark problem involving all the complexities of a BWR fuel assembly. The agreement of SUPERB results with both Monte Carlo and LWR-WIMS results is found to be excellent. (auth.)

  8. The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

    International Nuclear Information System (INIS)

    It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool keff is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor keff. In the interval in which the pitch was changed, keff decreased for up to ∼0.4 for standard and ∼0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, keff values are smaller for ∼0.2 (∼0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is ∼0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the keff of spent-fuel pool and can be neglected in spent-fuel pool design

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  10. The influence of pitch, burnup and absorber rods on the spent fuel pool criticality

    International Nuclear Information System (INIS)

    It has been shown that supercriticality might occur for some postulated accidents for the TRIGA spent fuel pool at ''Josef Stefan'' Institute in Ljubljana, Slovenia. However, in the previous studies, the effect of burnup was not accounted for. In this work the dependence of criticality on fuel burnup, the pitch among the elements and the number of uniformly mixed absorber rods for a square arrangement is presented. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. WIMS-D code was used to model the isotopic composition of the fuel for 5, 10, 20 and 30 % burnup without cooling time. The results show, that out of the three studied effects: pitch from contact (3.75cm) up to rack design pitch (8cm), number of absorbers from 0 to 8 and burnup up to 30 %, the pitch has the greatest influence on the multiplication factor keff. In the interval in which the pitch was changed, keff decreased for up to 0.45. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g. for contact of fuel elements with 8 absorber rods among them, keff values are smaller for almost 0.20 than for arrangement without absorber rods regardless of the burnup. The effect of burnup is the smallest since in no case keff decreases for more than 0.10, even for high burnups of 30 %. (author)

  11. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  12. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  13. Determination of the burn-up of TRIGA fuel elements by calculation and reactivity experiments

    International Nuclear Information System (INIS)

    The burnup of 17 fuel elements of the TRIGA Mark-II reactor in Vienna was measured. Different types of fuel elements had been simultaneously used for several years. The measured burnup values are compared with those calculated on the basis of core configuration and reactor operation history records since the beginning of operation. A one-dimensional, two-group diffusion computer code TRIGAP was used for the calculations. Comparison with burnup values determined by γ-scanning is also made. (orig./HP)

  14. Modeling of WWER-440 fuel pin behavior at extended burn-up

    Energy Technology Data Exchange (ETDEWEB)

    El-Koliel, Moustafa S. E-mail: moustafa_elkoliel@yahoo.com; Abou-Zaid, Attya A.; El-Kafas, A.A

    2004-04-01

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the 'rim effect'. High burn-up phenomena in WWER-440 UO{sub 2} fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO{sub 2} fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.

  15. Modeling of WWER-440 fuel pin behavior at extended burn-up

    International Nuclear Information System (INIS)

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the 'rim effect'. High burn-up phenomena in WWER-440 UO2 fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO2 fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

  16. Development of a fuel rod thermal-mechanical analysis code for high burnup fuel

    International Nuclear Information System (INIS)

    The thermal-mechanical analysis code for high burnup BWR fuel rod has been developed by NFI. The irradiation data accumulated up to the assembly burnup of 55 GWd/t in commercial BWRs were adopted for the modeling. In the code, pellet thermal conductivity degradation with burnup progress was considered. Effects of the soluble FPs, irradiation defects and porosity increase due to RIM effect were taken into the model. In addition to the pellet thermal conductivity degradation, the pellet swelling due to the RIM porosity was studied. The modeling for the high burnup effects was also carried out for (U, Gd)O2 and MOX fuel. The thermal conductivities of all pellet types, UO2, (U, Gd)O2 and (U, Pu)O2 pellets, are expressed by the same form of equation with individual coefficient γ in the code. The pellet center temperature was calculated using this modeling code, and compared with measured values for the code verification. The pellet center temperature calculated using the thermal conductivity degradation model agreed well with the measured values within ±150 deg. C. The influence of rim porosity on pellet center temperature is small, and the temperature increase in only 30 deg. C at 75 GWd/t and 200 W/cm. The pellet center temperature of MOX fuel was also calculated, and it was found that the pellet center temperature of MOX fuel with 10wt% PuO2 is about 60 deg. C higher than UO2 fuel at 75 GWd/t and 200 W/cm. (author)

  17. RAPID program to predict radial power and burnup distribution of UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Song, Jae Sung; Bang, Je Gun; Kim, Dae Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    Due to the radial variation of the neutron flux and its energy spectrum inside UO{sub 2} fuel, the fission density and fissile isotope production rates are varied radially in the pellet, and it becomes necessary to know the accurate radial power and burnup variation to predict the high burnup fuel behavior such as rim effects. Therefore, to predict the radial distribution of power, burnup and fissionable nuclide densities in the pellet with the burnup and U-235 enrichment, RAPID(RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet) program was developed. It considers the specific radial variation of the neutron reaction of the nuclides while the constant radial variation of neutron reaction except neutron absorption of U-238 regardless of the nuclides, the burnup and U-235 enrichment is assumed in TUBRNP model which is recognized as the one of the most reliable models. Therefore, it is expected that RAPID may be more accurate than TUBRNP, specially at high burnup region. RAPID is based upon and validated by the detailed reactor physics code, HELIOS which is one of few codes that can calculates the radial variations of the nuclides inside the pellet. Comparison of RAPID prediction with the measured data of the irradiated fuels showed very good agreement. RAPID can be used to calculate the local variations of the fissionable nuclide concentrations as well as the local power and burnup inside that pellet as a function of the burnup up to 10 w/o U-235 enrichment and 150 MWD/kgU burnup under the LWR environment. (author). 8 refs., 50 figs., 1 tab.

  18. Failure thresholds of high burnup BWR fuel rods under RIA conditions

    International Nuclear Information System (INIS)

    Transient deformation of high burnup boiling water reactor (BWR) fuel rods was measured and failure limit was examined under simulated reactivity-initiated accident (RIA) conditions. Brittle cladding failure occurred at a small hoop strain of about 0.4% during an early phase of the pulse irradiation tests at the Nuclear Safety Research Reactor (NSRR). Strain rates were in an order of tens %/s at the time of the failure. Comparison of the results with thermal expansion of pellets suggested that the cladding deformation was caused by thermal expansion of the pellets. In other words, the influence of fission gases in the pellets was small in the early phase deformation. Separate effect tests were conducted to examine influence of the cladding temperature on the cladding failure behavior. Influence of the pulse width on the failure threshold was discussed in terms of the strain rate, magnitude of the deformation and temperature of the cladding for high burnup BWR fuel rods under the RIA conditions. (author)

  19. Serviceability of VVER-1000 fuel rods at extended burn-up

    International Nuclear Information System (INIS)

    To-day in Russia fuel cycles are under development intended for fuel assembly burn-up up to 65MW.day/kgU. To validate the fuel serviceability standard VVER-1000 fuel assemblies are in trial operation to reach the burn-up >60MW.day/kgU. Late in 2002 fuel assemblies that reached the burn-up of 55MW.day/kgU after 5 years of operation were discharged from the Balakovo NPP. A series of assemblies will continue their trial operation during the 6th year. The post-irradiation examinations of high burn-up UO2 base fuel rods of the VVER type reveal that one of the most important consequences of burn-up extension is the so-called rim-effect. The term 'rim-effect' assumes a complex of specific features of the structure and thermophysical behaviour of a fuel rod related to processes proceeding in a fuel layer that is on the periphery of a fuel pellet

  20. Preliminary TRIGA fuel burn-up evaluation by means of Monte Carlo code and computation based on total energy released during reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Borio Di Tigliole, A.; Bruni, J.; Panza, F. [Dept. of Nuclear and Theoretical Physics, Univ. of Pavia, 27100 Pavia (Italy); Italian National Inst. of Nuclear Physics INFN, Section of Pavia, Via A. Bassi, 6, 27100 Pavia (Italy); Alloni, D.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Prata, M.; Salvini, A. [Italian National Inst. of Nuclear Physics INFN, Section of Pavia, Via A. Bassi, 6, 27100 Pavia (Italy); Applied Nuclear Energy Laboratory LENA, Univ. of Pavia, Via Aselli, 41, 27100 Pavia (Italy); Chiesa, D.; Clemenza, M.; Pattavina, L.; Previtali, E.; Sisti, M. [Physics Dept. G. Occhialini, Univ. of Milano Bicocca, 20126 Milano (Italy); Italian National Inst. of Nuclear Physics INFN, Section of Milano Bicocca, P.zza della Scienza, 3, 20126 Milano (Italy); Cammi, A. [Italian National Inst. of Nuclear Physics INFN, Section of Milano Bicocca, P.zza della Scienza, 3, 20126 Milano (Italy); Dept. of Energy Enrico Fermi Centre for Nuclear Studies CeSNEF, Polytechnic Univ. of Milan, Via U. Bassi, 34/3, 20100 Milano (Italy)

    2012-07-01

    Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the Univ. of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation. (authors)

  1. Preliminary TRIGA fuel burn-up evaluation by means of Monte Carlo code and computation based on total energy released during reactor operation

    International Nuclear Information System (INIS)

    Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the Univ. of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation. (authors)

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

    OpenAIRE

    Gholamzadeh Zohreh; Hossein Feghhi Seyed Amir; Soltani Leila; Rezazadeh Marzieh; Tenreiro Claudio; Joharifard Mahdi

    2014-01-01

    Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. N...

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

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

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

  4. IAEA Consultancy on Technical Influence of High Burnup UOX and MOX LWR Fuel on Spent Fuel Management

    International Nuclear Information System (INIS)

    This paper reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on SFM. These data will provide information on the impacts, regarding SFM, for those countries operating light-water reactors (LWR) and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. The effects of these property changes on wet and dry storage, transportation, reprocessing, refabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. The advantages and drawbacks of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, refabrication, and disposal will have to be balanced with the benefits of their use. (author)

  5. Effect of spent fuel burnup and composition on alteration of the U(Pu)O2 matrix

    International Nuclear Information System (INIS)

    For a potential performance assessment of direct disposal of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified. Leaching experiments were performed with five spent fuel samples from French power reactors (four UO2 fuel samples with burnup ratings of 22, 37, 47 and 60 GWd.THM-1 and a MOX fuel sample irradiated to 47 GWd.THM-1) to determine the release kinetics of the matrix containing most (over 95%) of the radionuclides. The experiments were carried out with granitic groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25 deg C) in a hot cell. After 1000 or 2000 days of leaching, the Sr/U congruence ratios for all the UO2 fuel samples ranged from 1 to 2, allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO2 fuel matrix. The daily strontium release factor was approximately 1 x 10-7 d-1 for UO2 fuel, and five to six times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings. Copyright (2001) Material Research Society

  6. 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

    International Nuclear Information System (INIS)

    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 keff 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 keff result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  7. Development plan of high burnup fuel for high temperature gas-cooled reactors in future

    International Nuclear Information System (INIS)

    Plan and status of research and development (R and D) were described on coated fuel particle (CFP) and fuel compacts for the core of small-sized high-temperature gas-cooled reactor (HTGR) HTR50S at second step of phase I (second core of HTR50S). Specifications of existing CFPs for high burnup (HTR50S2-type-CFPs) were adopted as specifications of CFPs, to reduce R and D. HTR50S2-type-CFPs were fabricated based on technology developed in High Temperature Engineering Test Reactor (HTTR) project. The first irradiation test of HTR50S2-type-CFPs is now being carried out. In addition, R and D for fuel compact with high packing fraction is planned, because volume fraction of UO2 kernel to whole of HTR50S2-type-CFP is rather smaller than that of the HTTR-type-CFP.We would aim to complete the proof of nuclear/thermal design of second core of HTR50S on integrity of fuel, feeding back results of R and Ds and nuclear/thermal design to each other. In addition, we describe outline of R and D plans for core of HTR50S in phase II and practical HTGR in Japan in future, naturally safe HTGR. (author)

  8. High burnup performance of Mg, Mg-Nb and Ti doped UO2 fuels

    International Nuclear Information System (INIS)

    In order to control irradiation performance of fuel swelling and FP gas release etc. at high burnups of light water reactor fuels, doped UO2 pellet fuels were prepared and their irradiation behavior was examined. The UO2 pellets doped 2.5 to 15mol%Mg, 5mol%Mg - 5mol%Nb, and 3.5mol%Ti and undoped UO2 pellets as a reference fuel were loaded together in a capsule and irradiated to the maximum burnups of 94GWd/t(U) below temperature of 1000degC in the JRR-3M reactor of JAERI. As results of post-irradiation examinations such as visual inspection, dimensional and density change measurements, thermal diffusivity and ceramography with optical microscope and EPMA, no difference was observed between the doped and the reference UO2 fuels. And valuable results were obtained on high burnup properties for swelling rates, thermal conductivities, structure changes and so on. (author)

  9. Room temperature leaching of labile radioactivity from irradiated PWR fuel according to the burnup

    International Nuclear Information System (INIS)

    Three PWR UO2 spent fuel specimens with average burnup of 22, 37 and 47 GWd tU-1 were submitted to sequential mode leaching in granitic groundwater for 62 cumulative days. The leaching rate decreased versus increasing contact time from 10-3 d-1 to 10-5 d-1. The 90Sr release appeared to be independent of the burnup with rates 2 orders of magnitude lower than for Cs but higher than the U and Pu release rates; both of the latter elements reached saturation rapidly, giving concentration values of 50-800 ppb and 0.1-10 ppb respectively, irrespective of the burnup. (authors)

  10. Development of high-burnup fuel analysis code EXBURN-I

    International Nuclear Information System (INIS)

    A computer code EXBURN-I has been developed which analyses LWR fuel behavior in high-burnup region in normal operation and transient conditions. In the high-burnup region, fuel behavior is affected considerably by such burnup-dependent factors as FP gas release, waterside corrosion of cladding, and pellet property change. To analyze these phenomena, in the present version, the base code FEMAXI-IV has been improved and incorporated such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding waterside corrosion. The present report describes the whole structure of the code, adopted models, and material properties, followed by input manual and sample input/output. Verification and further improvement of the code performance by experimental data will be done in the next stage. (author)

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

    International Nuclear Information System (INIS)

    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

  13. ANS/ENS tutorial session: Burnup credit issues in spent fuel transportation: Overview and objectives

    International Nuclear Information System (INIS)

    A number of opportunities exist to increase the efficiency of the next generation of spent fuel shipping casks. Improving cask efficiency will not only reduce life cycle transportation costs, but also is consistent with maintaining public and occupational radiological risks and, more importantly, total risks (radiological and nonradiological) within the guidelines of the ''as low as reasonably achievable'' (ALARA) philosophy. Increases in cask capacities will reduce both the total number of shipments required to transport a given amount of fuel and the number of handling operations at both shipping and receiving facilities. Additional capacity increases can be achieved by implementing various design strategies based on new concepts and/or the actual characteristics of the majority of the spent fuel to be shipped in the future. For example, it has been determined that additional capacity increases can be achieved by taking credit for burnup, the reduced reactivity that results when fuel has been used to produce power in a nuclear reactor. That is, as the fuel is used the atoms of fissile material decrease, and neutron absorbers (or ''poisons'') that tend to retard the fission process are produced. 7 refs., 1 fig

  14. Practices and developments in spent fuel burnup credit applications. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency convened a technical committee Meeting on Requirements, Practices and Developments in Burnup Credit (BUC) Applications in Madrid, Spain, from 22 to 26 April 2002. The purpose of this meeting was to explore the progress and status of international activities related to the BUC applications for spent nuclear fuel. This meeting was the third major meeting on the uses of BUC for spent fuel management systems held since the IAEA began to monitor the uses of BUC in spent fuel management systems in 1997. The first major meeting was an Advisory Group meeting (AGM), which was held in Vienna, in October 1997. The second major meeting was a technical committee meeting (TCM), which was held in Vienna, in July 2000. Several consultants meetings were held since 1997 to advise and assist the IAEA in planning and conducting its BUC activities. The proceedings of the 1997 AGM were published as IAEA-TECDOC-1013, and the proceedings of the 2000 TCM as IAEA-TECDOC-1241. BUC for wet and dry storage systems, spent fuel transport, reprocessing and final disposal is needed in many Member States to allow for increased enrichment, and to increase storage capacities, cask capacities and dissolver capacities avoiding the need for extensive modifications. The use of BUC is a necessity for spent fuel disposal

  15. Fuel rod and core materials investigations related to LWR extended burnup operation

    Science.gov (United States)

    Kolstad, Erik; Vitanza, Carlo

    1992-06-01

    The paper deals with tests and recent measurements related to extended burnup fuel performance and describes test facilities and results in the areas of waterside cladding corrosion and irradiation-assisted stress corrosion cracking (IASCC). Fuel temperature data suggest a gradual degradation of UO 2 thermal conductivity with exposure in the range 6-8% per 10 MWd/kgUO 2 at temperatures below 700°C. The effect on the fuel microstructure of interlinkage and resintering phenomena is shown by measuring the surface-to-volume ( S/ V) ratio of the fuel. Changes in S/V with burnup are correlated to power rating and fuel operating temperature. No evidence was found of enhanced fission gas release during load-follow operation in the burnup range 25-45 MWd/kgUO 2. The effect of high lithium concentration (high pH) on the corrosion behaviour of pre-irradiated high burnup Zircaloy-4 fuel rods subjected either to nucleate boiling or to one-phase cooling conditions was studied. The oxide thickness growth rates measured at an average burnup up to 40 MWd/kgUO 2 are consistent with literature data and show no evidence of corrosion enhancement due to the high lithium content and little effect of cooling regime. A test facility for exploring the effects of environmental variables on IASCC behaviour of in-core structural materials is described.

  16. High burnup fuel simulation during a reactivity initiated transient using the FRAPTRAN code

    International Nuclear Information System (INIS)

    CIP01 test simulated a RIA (Reactivity Initiated Accident) transient at the experimental CABRI reactor under sodium cooling conditions on November 29th 2001. The father rod was fabricated by ENUSA and irradiated during five cycles in Vandellos 2 NPP and it reached a maximum burnup at EOL (End of Life) of 75 GWd/tU. The test rodlet was refabricated in Studsvik from the span 5 of the father rod. The cladding material was ZIRLO with a mean oxide layer of 75 μm. This note describes the modelling of the CIP01 test carried out in the CABRI reactor under RIA conditions by the use of the transient FRAPTRAN code (1.2 version). The CIP01 test consisted in a power pulse of 32.4 ms with a deposited energy of 98.2 cal/g at PPN (Peak Power Node). The code results will be presented and compared to the experimental data available. In addition, the FRAPTRAN and the SCANAIR code estimations will be also contrasted. This work is framed within the CSN-CIEMAT agreement on 'Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup'. (Author)

  17. High burnup fast reactor fuel: processing and waste management experiences

    International Nuclear Information System (INIS)

    The routine processing of mixed Plutonium/Uranium oxide fuels from the Prototype Fast Reactor (PFR) at Dounreay began in September 1980 and the design features of the modified Dounreay Fast Reactor (DFR) reprocessing plant and experience of the first active campaign were described in a paper to the British Nuclear Engineering Society in November 1981 (1). Since then progress in processing the fuel discharged from PFR has been covered briefly in a number of papers to international conferences and the Public Inquiry held in 1986 into the outline planning application for the proposed European Demonstration Reprocessing Plant. During this decade considerable experience in the operation of fast reactors and associated fuel plants has been accumulated providing confidence in the system before entering the next development phase - that of its commercial demonstration. Confidence in the UK draws on the successful operation of the PFR and the associated Dounreay fuel reprocessing and BNF Sellafield fabrication plants. Of equal importance is public confidence in safe operation and in the management of wastes generated by a fast reactor system. The present paper is a review of fast reactor reprocessing and waste management at the Dounreay Nuclear Establishment (DNE) as a contribution to the present status of the fast reactor system

  18. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  19. Self-organized criticality in evolution of nuclear fuel microstructure

    International Nuclear Information System (INIS)

    Nuclear fuel microstructure has major influence on the fission product release from nuclear fuel matrix. Here we present the self-organized criticality model applied to describe the evolution of nuclear fuel microstructure. It is shown that the behavior of fuel bubbles is similar to that of species in natural ecosystems and their evolution can be characterized as an avalanche process. Modelled bubble size distribution for different fuel burnups is in good agreement with the experimental data. (author)

  20. Criticality safety analysis of WWER-440 spent fuel cask with radial and axial burnup profile implementation

    International Nuclear Information System (INIS)

    The impact of radial and axial burnup profile on the criticality of WWER-440 spent fuel cask is presented in the paper. The calculations are performed based on two AER Benchmark problems for WWER-440 irradiated fuel assembly. The radial zonewise dependent spent fuel inventory has been calculated by the NESSEL - NUKO code system. The axial dependent isotope concentrations have been determined by the modular code system SCALE4.4. For criticality calculations the SCALE4.4 has been applied. Calculations have been carried out for cask with 30 WWER-440 fuel assemblies with initial enrichment 3.6% of 235U and burnup up to 40 MWd/kgU. The influence of radial and axial burnup credit on the cask criticality has been evaluated

  1. Irradiation behavior of FBTR mixed carbide fuel at various burn-ups

    International Nuclear Information System (INIS)

    The fast breeder test reactor at Kalpakkam has completed nearly 25 years of operation and is now operating at 18 MWt capacity with 46 fuel subassemblies (FSA) in the core consisting of 27 Mark-I (70% PuC + 30% UC), 13 Mark-II (55% PuC + 45% UC) and 6 MOX (44% PuO2 + 56% UO2) and one test PFBR FSA. Post Irradiation Examination (PIE) campaigns on FSAs at different burnup levels has provided valuable information about the irradiation behavior of the carbide fuel. This paper gives a summary of the irradiation performance of the carbide fuel evaluated through some of the investigations such as neutron radiography, x-radiography, gamma scanning, fission gas analysis and ceramography. Burnup of the carbide fuel could be enhanced from the initial design burnup limit of 50 GWd/t to 165 GWd/through systematic PIE. (author)

  2. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    Energy Technology Data Exchange (ETDEWEB)

    Ozdemir, Levent, E-mail: levent.ozdemir@taek.gov.tr [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey); Acar, Banu Bulut; Zabunoglu, Okan H. [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey)

    2011-02-15

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of {sup 239}Pu and {sup 241}Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  3. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W. [OECD Halden Reactor Project (Norway)

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  4. Spent fuel dissolution test including dissolver off-gas treatment in the NUCEF (Nuclear Fuel Cycle Safety Engineering Facility) alpha-gamma cell: Spent fuels with burnups up to 44 GWDT-1

    International Nuclear Information System (INIS)

    Spent fuel dissolution test was carried out to study the dissolution behavior of nuclides and the behavior of volatile nuclides, using a bench-scale reprocessing test rig in the NUCEF alpha-gamma cell. 29 and 44 GWdt-1 spen fuels were respectively dissolved. Study on the dissolution behavior showed that the dissolution rate of uranium and plutonium were similar. Also, other species: Cesium, Americium, Strontium and Neodymium were found to dissolve in the similar manor to uranium. The fraction of Mo dissolved was found to decrease during the dissolution. Zirconium molybdate (Zr(Mo2O7)(OH)2(H2O)2) was identified from the XRD pattern of residue in the 29 GWdt-1 spent fuel dissolution. Those results suggested the precipitation of the hydrate during dissolution. As for the iodine-129 removal, decontamination factor of AGS (silica-gel impregnated with silver nitrate) column was more than 36,000. Measurement of iodine-129 in the dissolver solution revealed that less than 0.57% of total iodine-129 generated, which was estimated by ORIGEN II calculation, was remained in the dissolver solution. Also, measurement of iodine-129 by an iodine-stripping operation from dissolver solution using potassium iodate suggested that another 2.72% of total iodine-129 precipitated as iodide. In addition, about 70% of total iodine generated was measured in the AGS columns. Rest of iodine-129 was supposed to adsorb to the HEPA filter and the inner surface of dissolver off-gas lines. Those results on iodine-129 distribution were found to be almost identical to the results obtained in the study using iodine-131 as tracer. Release of carbon-14 as carbon dioxide during dissolution was found to occur when the release of Kr-85. From the measurement of carbon-14, initial nitrogen-14 concentration in the fuel was estimated to be about several ppms, which was within the range reported. (author)

  5. Fission gas release and pellet microstructure change of high burnup BWR fuel

    International Nuclear Information System (INIS)

    UO2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

  6. Experimental support of WWER-440 fuel reliability and serviceability at high burnup

    International Nuclear Information System (INIS)

    Results from post-reactor examination of two WWER-440 fuel assemblies spent at the Kola NPP Unit 3 during 4 and 5 fuel cycles are presented. The fuel assembly states and their serviceability allowance are estimated experimentally at the RIAR hot laboratory and studied by non-destructive and destructive methods. The following parameters are examined: fuel assembly overall dimensions change; fuel element diameter change; fuel element cladding corrosion and hydriding; fuel element cladding mechanical properties; fission gas release from fuel and gas pressure; fuel macro- and microstructure. it has been found that the maximum fuel burnup of fuel assemblies No. 1 and No.2 achieved is 58.3 and 64.0 MWd/kg, respectively. The mechanical fuel pellets-cladding interaction has been observed at the average fuel burnup above 45 MWd/kg that occurred with increasing the local cladding diameter at the areas of pellets end arrangement (bamboo stick). The gas release linearly increases at the range 2.7% per 10 MWd/kg within burnup of 43-60 MWd/kg. 9 figs., 3 refs

  7. Impact of axial burnup profile on criticality safety of ANPP spent fuel cask

    International Nuclear Information System (INIS)

    Criticality safety assessment for WWER-440 NUHOMS cask with spent nuclear fuel from Armenian NPP has been performed. The cask was designed in such way that the neutron multiplication factor keff must be below 0,95 for all operational modes and accident conditions. Usually for criticality analysis, fresh fuel approach with the highest enrichment is taken as conservative assumption as it was done for ANPP. NRSC ANRA in order to improve future fuel storage efficiency initiated research with taking into account burn up credit in the criticality safety assessment. Axial burn up profile (end effect) has essential impact on criticality safety justification analysis. However this phenomenon was not taken into account in the Safety Analysis Report of NUHOMS spent fuel storage constructed on the site of ANPP. Although ANRA does not yet accept burn up credit approach for ANPP spent fuel storage, assessment of impact of axial burnup profile on criticality of spent fuel assemblies has important value for future activities of ANRA. This paper presents results of criticality calculations of spent fuel assemblies with axial burn up profile. Horizontal burn up profile isn't taken account since influence of the horizontal variation of the burn up is much less than the axial variation. The actinides and actinides + fission products approach are discussed. The calculations were carried out with STARBUCS module of SCALE 5.0 code package developed at Oak Ridge National laboratory. SCALE5.0 sequence CSAS26 (KENO-VI) was used for evaluation the keff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect

  8. Fuels for Advanced Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Fuels for advanced nuclear reactors differ greatly from conventional light water reactor fuels and vary widely between the different concepts, due differences in reactor architecture and deployment. Functional requirements of all fuel designs include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintaining a coolable geometry. In all cases, the anticipated fuel performance under normal or off-normal conditions is the limiting factor in reactor system design, and cumulative effects of increased exposure to higher burnup degrades fuel performance. In high-temperature (thermal) gas reactor systems, fuel particles of uranium dioxide or uranium oxycarbide particles are coated with layers of carbon and SiC (or ZrC). Such fuels have been used successfully to very high burnup (10-20% of heavy-metal atoms) and can withstand transient temperatures up to 1600 C. Oxide (pellet-type) and metal (pin-type) fuels clad in stainless steel tubes have been successfully used in liquid metal cooled fast reactors, attaining burnup of 20% or more of heavy-metal atoms. Those fuel designs are being adapted for actinide management missions, requiring greater contents of minor actinides (e.g. Am, Np, Cm). The current status of each fuel system is reviewed and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed

  9. Development of the CANDU high-burnup fuel design/analysis technology

    International Nuclear Information System (INIS)

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

  10. Investigation of research and development subjects for the Very High Burnup Fuel

    International Nuclear Information System (INIS)

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)

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

    International Nuclear Information System (INIS)

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

  12. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science

    2014-12-15

    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  13. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    International Nuclear Information System (INIS)

    This paper aims to investigate 232Th/233U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. 232Th/235U/238U oxide mixture was considered as fuel in the core, when the mass fraction of 232Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of 238U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the 232Th, 233U, 238U, 237Np, 239Pu, 241Am and 244Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  14. Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

    International Nuclear Information System (INIS)

    Highlights: • A 300 MWt Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

  15. Impact of fission gas on irradiated PWR fuel behaviour at extended burnup under RIA conditions

    International Nuclear Information System (INIS)

    With the world-wide trend to increase the fuel burnup at discharge of the LWRs, the reliability of high burnup fuel must be proven, including its behaviour under energetic transient conditions, and in particular during RIAs. Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup. The potential for swelling and transient expansion work under rapid heating conditions characterizes the high burnup fuel behaviour by comparison to fresh fuel. This effect is resulting from the steadily increasing amount of gaseous and volatile fission products retained inside the fuel structure. An attempt is presented to quantify the gas behaviour which is motivated by the results from the global tests both in CABRI and in NSRR. A coherent understanding of specific results, either transient release or post transient residual retention has been reached. The early failure of REP Na1 with consideration given to the satisfactory behaviour of the father rod of the test pin at the end of the irradiation (under load follow conditions) is to be explained both by the transient loading from gas driven fuel swelling and from the reduced clad resistance due to hydriding. (R.P.)

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

    International Nuclear Information System (INIS)

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

  17. Determination of burn-up of irradiated PHWR fuel samples from KAPS-1 by mass spectrometry

    International Nuclear Information System (INIS)

    Burn-up was determined experimentally using thermal ionization mass spectrometry for three spent UO2 fuel samples, which had undergone extended irradiation in Kakrapar Atomic Power Station Unit 1 (KAPS-1). The method involves dissolution of the irradiated fuel sample, separation and determination of burn-up monitor, uranium and plutonium. Isotope Dilution-Thermal Ionisation Mass Spectrometry (ID-TIMS) using Triple Spike Mixture consisting of (142Nd+233U+242Pu) was employed for the concentration determination of Nd, U and Pu in the dissolved fuel samples. The atom percent fission was calculated based on 148Nd as a burn-up monitor and also from the changes in the abundances of heavy element isotopes. Fractional fission contributions from the major fissile nuclides were calculated from heavy elemental data and also from the Nd isotopic ratios. (author)

  18. Effect of the Lattice Pitch Variation on the Burn-up of the ETRR-1 Reactor Fuel

    International Nuclear Information System (INIS)

    The recent development in the field of reactor physics and nuclear safety calculation is providing a modern aspect to the fuel cycle economics through the reactor operations. For that, it is necessary to review the previous design of the old operating reactors. One of these reactors is the ETRR-1, which is operating since 1961. The fuel basket of this reactor contains 16 fuel rods arranged in 4 x 4 shape with a lattice pitch 17.5 mm. The lattice cell parameters and burn-up calculations were performed by using the WIMSD4 code for different lattice pitch configurations including the exist one. This study shows that, the present lattice pitch of the ETRR-1 reactor fuel element is not the best one and the 15 mm lattice pitch is more convenient from the fuel cycle economics point of view

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

    International Nuclear Information System (INIS)

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B4C 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, keff, 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 keff 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, keff 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

  20. Determination of the fuel element burn-up for mixed TRIGA core by measurement and calculation with new TRIGLAV code

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)

  1. Fission gas release behavior in high burnup UO2 fuels with developed rim-structure

    International Nuclear Information System (INIS)

    The effect of rim structure formation and external restraint pressure on fission gas release at transient conditions has been examined by using an out-of-pile high pressure heating technique for high burnup UO2 fuels (60, 74 and 90 GWd/t), which had been irradiated in test reactors. The latter two fuels bore a developed rim structure. The maximum heating temperature was 1500 degC, and the external pressures were independently controlled in the range of 10-150 MPa. The present high burnup fuel data were compared with those of previously studied BWR fuels of 37 and 54 GWd/t with almost no rim structure. The fission gas release and bubble swelling due to the growth of grain boundary bubbles and coarsened rim bubbles were effectively suppressed by the strong restraint pressure of 150 MPa for all the fuels; however the fission gas release remarkably increased for the two high burnup fuels with the developed rim structure, even at the strong restraint conditions. From the stepwise de-pressurization tests at an isothermal condition of 1500degC, the critical external pressure, below which a large burst release due to the rapid growth and interlinkage of the bubbles abruptly begins, was increased from a 40-60 MPa level for the middle burnup fuels to a high level of 120-140 MPa for the rim-structured high burnup fuels. The high potential for transient fission gas release and bubble swelling in the rim-structured fuels was attributed to highly over-pressurized fission gases in the rim bubbles. (author)

  2. The use of burnup credit in criticality control for the Korean spent fuel management program

    International Nuclear Information System (INIS)

    More than 25% k-eff saving effect is observed in this burnup credit analysis. This mainly comes from the adoption of actinide nuclides and fission products in the criticality analysis. By taking burnup credit, the high capacity of the storage and transportation can be more fully utilized, reducing the space of storage and the number of shipments. Larger storage and fewer shipments for a given inventory of spent fuel result should in remarkable cost savings and more importantly reduce the risks to the public and occupational workers for the Korean Spent Fuel Management Program

  3. 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.

  4. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  5. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA's activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ''FUMEX'', allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs

  6. Nondestructive methods of determination of isotope composition and burnup of spent fuel from WWER-type reeactor

    International Nuclear Information System (INIS)

    Application of the nondestructive methods of analysis (NDA) is discussed being used for determination of burnup and isotope ratio as well as composition of spent fuel elements in the fuel assemblies of WWER-type reactor. Results are discussed which have been obtained by means of the γ-spectrometric method. Prospects are noted of a semiempiric method of determination of the burnup and isotope ratio of the fuel in WWER-type reactors. This method is based on the combination of the data which have been obtained by the γ-spectroscopy and of calculation. NDA has been considered which is based on the registration of self radiation of neutrons from spent fuel elements and assemblies. This method has some advantages as to compare to the γ-spectrometric one which permits to hope for successful, in the case of it's firther elaboration, appliaction of the neutron passive method both in the fuel cycle and in the safeguard system. In the conclusion, prospects are discussed of development and application of the NDA for spent fuels from nuclear power plants

  7. Feasibility and incentives for burnup credit in spent-fuel casks

    International Nuclear Information System (INIS)

    The spent-fuel carrying capacities of previous-generation spent-fuel shipping casks have been primarily thermal and/or shielding limited. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced considerably and cask capacities become criticality limited. Using burnup credit in the design of future casks can result in increased cask capacities as well as reduced environmental impacts and savings in time and money

  8. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  9. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    International Nuclear Information System (INIS)

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am242m is Am241(n,γ-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am242m concentration.The uncertainties in the inventory of fission products reach up to 30%

  10. Fuel burnup calculation for HEU and LEU cores of Ghana MNSR

    International Nuclear Information System (INIS)

    Fuel burnup calculations have been performed using a computer program developed as part of this research work for both Highly Enriched Uranium (90.2 % U-235) and Low Enriched Uranium (12.6 % U-235) cores for Ghana Research Reactor-1 (GHARR-1). Fuel depletion analyses of the GHARR-1 core was also performed which provided an inventory of the actinides formed as a result of burnup. The effect of the production of plutonium isotopes with burnup on reactor operation was also estimated. A FORTRAN 95 code was written based on the three group model approach namely fast, resonance and slow (thermal) neutron reactions. The time rate of change of each fuel isotope density is given by a first order differential equation. A general solution for each fuel isotope rate equation was used as input for the computer code. These results are particularized to the case of constant power during a short time interval, during which the slow (thermal) neutron flux is considered constant. The results obtained for the HEU were in good agreement with those found in literature. Therefore, this code can be used to estimate the burnup of LEU fuel for core conversion from HEU to LEU. (au)

  11. Burn-up and cycle length optimization project of the robust fuel programme

    International Nuclear Information System (INIS)

    The Spanish electric sector (UNESA) takes part in the Robust Fuel programme in the different work groups set up by EPRI. Iberinco, with the collaboration of Iberdrola Generacion (TECNO and Cofrentes NPP) and Soluziona Ingenieria, has created a stable multidisciplinary group to assimilate and follow up this program, analyzing in detail the technology generated and evaluating the conclusions to provide the most suitable recommendations for application. Along these lines, one of the most promising projects within technical group 3 (High burn properties) has been the one called Burn-up and cycle Length Optimization. In January 2000 Duke Power published a study on the plants it owns (PWR type) and 18-month cycles, to establish the optimum unloading burn-up of fuel. The conclusion it reached is that the fuel cost drops t a minimum for average unload burn-ups of between 60 and 70 GWd/MTU. As an extension to this study and covering a wider base of considerations, Exelon, with the support of Westinghouse and the University of Pennsylvania, released a study in December 2001 on different reference cores with different cycle lengths. In this study, the optimum burn-up without exceeding current maximum enrichment limits (5%) is determined. Publication of the results of the second phase, considering higher enrichments, was due in the summer of 2002. The design of the core to be refueled and economic analyzes show that both pressurized water reactors (PWR) and boiling water reactors (BWR) can obtain significant benefits by increasing the fuel unloading burn-up above currently licensed limits. However, the optimum unload burn-up level is not reached without exceeding the current enrichment limit of 5% . (Author)

  12. 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.

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

    International Nuclear Information System (INIS)

    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

  14. Renewal plan of the JMTR hot laboratory for the irradiation test of high burn-up fuels in FY2008

    International Nuclear Information System (INIS)

    Refurbishment of the Japan Materials Testing Reactor (JMTR), which is recognized as one of important facilities in Japan for safety research, is in progress by the JAEA. In 'Extensive safety research of light-water reactor (LWR) fuels and materials' under a contract with the Nuclear and Industrial Safety Agency of Ministry of Economy, Trade and Industry of Japan, the irradiation tests are planned in order to examine integrity of the LWR fuels and structure materials. For the irradiation tests of high burnup fuels and irradiated materials in the JMTR, modification of the hot laboratory facilities are needed, which are (1) preparing of documents books for strengthening JMTR hot-lab. cell-shielding, (2) detailed design of the capsule assembling device, (3) safety analysis for domestic transportation cask, and (4) confirmation test of diamond drill of fuel-rod center-hole processing. This paper summarizes investigation results for the design and manufacture, future on these equipments conducted in FY2008. (author)

  15. Renewal plan of the JMTR hot laboratory for the irradiation test of high burn-up fuels in FY2007

    International Nuclear Information System (INIS)

    Refurbishment of the Japan Materials Testing Reactor (JMTR), which is recognized as one of important facilities in Japan for safety research, is in progress by the JAEA. In 'Extensive safety research of light-water reactor (LWR) fuels and materials' under a contract with the Nuclear and Industrial Safety Agency of Ministry of Economy, Trade and Industry of Japan, the irradiation tests are planned in order to examine integrity of the LWR fuels and structure materials. For the irradiation tests of high burnup fuels and irradiated materials in the JMTR, modifications of the hot laboratory facilities are needed, which are (1) strengthening JMTR hot-lab. cell-shielding, (2) the capsule assembling device, (3) domestic transportation cask, (4) fuel-rod center-hole processing device, (5) master-slave manipulators, (6) power manipulator, and (7) scanning electron microscope. In FY2007, above related modifications were investigated. This paper summarizes the investigation results of them. (author)

  16. Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations

    Science.gov (United States)

    Sjöstrand, H.; Alhassan, E.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.

    2014-04-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor lifetime using the Total Monte Carlo approach (TMC). Within the TENDL project, nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.

  17. Specific application of burnup credit for MOX PWR fuels in the rotary dissolver

    International Nuclear Information System (INIS)

    In prospect of a Mixed OXide spent fuels processing in the rotary dissolver in COGEMA/La Hague plant, it is interesting to quantify the criticality-safety margins from the burnup credit. Using the current production computer codes and considering a minimal fuel irradiation of 3 200 megawatt-day per ton, this paper shows the impact of burnup credit on industrial parameters such as the permissible concentration in the dissolution solution or the permissible oxide mass in the rotary dissolver. Moreover, the burnup credit is broken down into five sequences in order to quantify the contribution of fissile nuclides decrease and of minor actinides and fission products formation. The implementation of the burnup credit in the criticality-safety analysis of the rotary dissolver may lead to workable industrial conditions for the particular MOX fuel studied. It can eventually be noticed that minor actinides contribution is negligible and that considering only the six major fission products is sufficient, owing to the weak fuel irradiation contemplated. (author)

  18. Burn-Up Dependence of Bubble Morphology of Uranium Silicide Dispersion Fuels Used in Research Reactor

    International Nuclear Information System (INIS)

    Burn-up dependence of fission gas bubble morphology of U3Si2-Al and U3Si-Al dispersion fuels are reviewed with the data of ANL(Argonne Nation Laboratory) and KAERI(Korea Atomic Energy Research Institute

  19. Review of the IAEA nuclear fuel cycle and material section activities connected with nuclear fuel including WWER fuel

    International Nuclear Information System (INIS)

    Program activities on Nuclear Fuel Cycle and Materials cover the areas of: 1) raw materials (B.1.01); 2) fuel performance and technology (B.1.02); 3) pent fuel (B.1.03); 4) fuel cycle issues and information system (B.1.04); 5) support to technical cooperation activities (B.1.05). The IAEA activities in fuel performance and technology in 2001 include organization of the fuel experts meetings and completion of the Co-ordinate Research Projects (CRP). The special attention is given to the advanced post-irradiation examination techniques for water reactor fuel and fuel behavior under transients and LOCA conditions. An international research program on modeling of activity transfer in primary circuit of NPP is finalized in 2001. A new CRP on fuel modeling at extended burnup (FUMEX II) has planed to be carried out during the period 2002-2006. In the area of spent fuel management the implementation of burnup credit (BUC) in spent fuel management systems has motivated to be used in criticality safety applications, based on economic consideration. An overview of spent fuel storage policy accounting new fuel features as higher enrichment and final burnup, usage of MOX fuel and prolongation of the term of spent fuel storage is also given

  20. Present status and future developments of the implementation of burnup credit in spent fuel management systems in Germany

    International Nuclear Information System (INIS)

    This paper describes the experience gained in Germany in implementing burnup credit in wet storage and dry transport systems of spent PWR, BWR, and MOX fuel. It gives a survey of the levels of burnup credit presently used, the regulatory status and activities planned, the fuel depletion codes and criticality calculation codes employed, the verification methods used for validating these codes, the modeling assumptions made to ensure that the burnup credit criticality analysis is based on a fuel irradiation history which leads to bounding neutron multiplication factors, and the implementation of procedures used for fuel loading verification. (author)

  1. Development and preliminary analyses of material balance evaluation model in nuclear fuel cycle

    International Nuclear Information System (INIS)

    Material balance evaluation model in nuclear fuel cycle has been developed using ORIGEN-2 code as basic engine. This model has feature of: It can treat more than 1000 nuclides including minor actinides and fission products. It has flexibility of modeling and graph output using a engineering work station. I made preliminary calculation of LWR fuel high burnup effect (reloading fuel average burnup of 60 GWd/t) on nuclear fuel cycle. The preliminary calculation shows LWR fuel high burnup has much effect on Japanese Pu balance problem. (author)

  2. Comparisons of the predicted and measured isotopic composition for high burnup PWR spent fuels

    International Nuclear Information System (INIS)

    Comparisons between the calculated and measured isotopic composition for high burnup Korean PWR spent fuel samples were carried out. Spent fuel samples used in this study were obtained from commercial Korean PWRs, Ulchin unit 2 and Yonggwang unit 1. A radiochemical analysis of the spent fuel samples was performed to determine the isotopic compositions of U, Pu, and Nd. The depletion calculations which were carried out using the SAS2H control module in Version 5.1 of the SCALE code system were compared with the results of the radiochemical analyses. The results derived from the measured and calculated concentrations for each isotope of the corresponding samples were generally consistent with the earlier studies and the results were different within a few percent. The validity of the SAS2H control module in Version 5.1 of the SCALE code system could be confirmed in a high burnup spent fuel above 45 GWd/MTU

  3. TRIGA fuel burn-up calculations supported by gamma scanning

    International Nuclear Information System (INIS)

    High resolution gamma-ray spectroscopy based non-destructive methods is employed to measure spent fuel parameters. By this method, the axial distribution of Cesium-137 has been measured which results in an axial burn up profiles. Knowing the exact irradiation history of the fuel, four spent TRIGA fuel elements have been selected for on-site gamma scanning using a special shielded scanning device developed at the Atominstitute. Each selected fuel element was transferred into the fuel inspection unit using the standard fuel transfer cask. Each fuel element was scanned in one centimetre steps of its active fuel length and the Cesium-137 activity was determined as a proved burn up indicator. The absolute activity of each centimetre was measured and compared with the reactor physics code ORIGEN2.2 results. This code was used to calculate average burn up and isotopic composition of fuel element. The comparison between measured and calculated results shows good agreement. (author)

  4. Technologies for manufacturing UO2 sintered pellets to fuel burnup extension

    International Nuclear Information System (INIS)

    The actual tendency all over the world is to manufacture fuel bundles capable to resist high burn-up. The factors affecting the burn-up increase are: the pellet-cladding mechanical interaction (PCMI), the oxidation and hydriding of the Zircaloy-4 sheath, the increase of internal pressure, stress corrosion cracking, Zircaloy-4 irradiation growth, fuel swelling. A way to increase fuel burn-up is to diminish the elements internal pressure by adequate UO2 fuel pellet structure (large grain or controlled closed porosity). In the large grain size UO2 pellets, fission gas release rate decreases and the elements internal pressure increase slowly. Similarly, in the UO2 sintered pellet with controlled closed porosity the fission gas accommodation is better and the elements internal pressure increases slowly. The paper presents a literature review related to the technologies and the methods for manufacturing UO2 sintered pellets to fuel burn-up extension. The flowsheets for large grains and controlled closed porosity UO2 sintered pellets obtained by Nb2O5 dopant respectively pores former addition in UO2 sinterable powder, pressing and sintering in H2 atmosphere are exposed. In the diagrams are presented the dependency of the main sintered pellet characteristics (pore radius distribution, pores volume, density, grains size) as function of the Nb2O5 dopant concentration, UO2 sinterable powder nature and sintering temperature. Other sintered pellets characteristics (electrical conductivity, Seebeck coefficient, high temperature molar heat capacity and thermomechanical properties) are presented. The technologies for sintered pellets manufacturing for RU, DUPIC, MOX fuel cycles are presented. A proposal related to fuel manufacturing from Uranium compound resulted in LWR spent fuel reprocessing is also given. (author)

  5. A combined 1D/3D fuel burnup analysis of generation IV light water reactor IRIS

    International Nuclear Information System (INIS)

    A combined 1D/3D methodology for the fuel burnup analysis of generation IV light water reactors with thin boron coating that covers the fuel rods is described in this paper. This methodology is founded on three approximations. The first approximation assumes that the problem of fuel depletion in the entire 3D core can be resolved into two independent problems. One is a 3D Monte Carlo evolution of power distribution in large volumes (nodes) with the KENO-V.a code, and the other is a transport method evolution of burnup dependent fuel composition in 1D Wigner-Seitz cell for each node independently. With the second approximation, the time-dependent fuel composition in the node (e.g., in the fuel assembly) is calculated by using a 1D fuel depletion analysis with the SAS2H control module from the SCALE-4.4a code system. The third approximation involves smearing the boron coating with the clad (by volume homogenization). The proposed SAS2H/KENO-V.a methodology is verified for the case of 2D x-y model of IRIS 15x15 fuel assembly (with a reflective boundary condition) by using two well benchmarked code systems. The first one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. It has been found that the proposed SAS2H/KENO-V.a methodology gives a satisfactory accuracy for keff and nuclide composition. Finally, this methodology was applied for 3D burnup analysis of IRIS-1000 benchmark≠44 core. Detailed keff and power density evolution with burnup are reported. (author)

  6. Cogema and the nuclear fuel. A clue role in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The present issue of 'Les Cahiers de COGEMAGAZINE' addresses the topics of nuclear fuel production especially for PWR and Breeder Reactors. The papers deal with: the sketchy history of French nuclear industry, the economy and fuel marketing, the situation of the PWR programme, the fuels for breeder and research reactors. In the end prospective and concluding considerations are given. The most significant lines of progress related to the new fuels are estimated to be: high burn-up (by increasing the resistance to fission gas pressure and irradiation), improvement of response to power excursions, fuel matrices of stronger retention, increase in the plutonium content of MOF, 100% MOF-fuelled reactors, optimizing the utilization of consumable poisons (for PWR) and very high burn-up and very long service lifetimes (for breeders)

  7. 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)

  8. Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

    International Nuclear Information System (INIS)

    Highlights: ► Depletion isotopics are needed for burnup credit in spent fuel pool analyses. ► We developed ISOCRIT to generate the isotopics using conservative depletion assumptions. ► ISOCRIT works in an automated fashion passing data between lattice physics and 3D Monte Carlo codes. ► Analyses to assess the impact of different depletion parameters on the reactivity of the spent fuel in pool conditions. - Abstract: In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse’s state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.

  9. Propagation of nuclear data uncertainties for ELECTRA burn-up calculations

    CERN Document Server

    ostrand, H; Duan, J; Gustavsson, C; Koning, A; Pomp, S; Rochman, D; Osterlund, M

    2013-01-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in Pu-239 transport data to uncertainties in the fuel inventory of ELECTRA during the reactor life using the Total Monte Carlo approach (TMC). Within the TENDL project the nuclear models input parameters were randomized within their uncertainties and 740 Pu-239 nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty in the ...

  10. 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.

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

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    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...

  12. ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

    International Nuclear Information System (INIS)

    1 - Historical background and information: - 28-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. - 37-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the Candu Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency with reactor physics codes. 2 - Application of the data: The libraries in this data collection are designed for characterising used fuel from Candu pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. 3 - Source and scope of data: The Candu libraries are updated with cross sections from a variety of different sources. Capture

  13. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    International Nuclear Information System (INIS)

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (keff) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of keff. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of keff values calculated by the participants from the mean value is almost within the band of ±1%Δk/k. The deviations from the averaged calculated fission rate profiles are found to be within ±5% for most cases. (author)

  14. Measurements of fuel burnup for the RA reactor spent fuel elements stored in the stainless steel containers (Draft version)

    International Nuclear Information System (INIS)

    According to the Radiological Characterisation Plan of the RA reactor, the accurate data on fuel burnup are very important for the radiation safety provisions during removal of spent fuel elements from the RA reactor as well as for verification of methods, geometry models and historically reviewed data concerning fuel irradiation. These data and methods will be used for neutron flux calculations in the RA reactor cores, reflector and biological shield, and finally for activity calculations of hard-to-detect radionuclides in the graphite reflector and concrete shields. Since the comparison of previous experimental data with the calculations showed discrepancy of 25% , fuel burnup of all fuel elements stored in the stainless steel containers was measured recently (from february to August 2006). This progress report summarizes the techniques and methods used for fuel burnup measurements of both type fuel elements (2% enriched metal uranium and 80% enriched uranium dioxide). It presents results for some maximum burned fuel elements and contains results of multichannel scanning of gamma ray emission from all stainless steel containers with spent fuel elements in storage pool

  15. Fuel Modelling at Extended Burnup (Fumex-II). Report of a Coordinated Research Project 2002-2007

    International Nuclear Information System (INIS)

    It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This, in turn, leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and unproved operating economics. Reliable prediction of fuel behaviour constitutes a basic demand for safety based calculations, for design purposes and for fuel performance assessments. Owing to the large number of interacting physical, chemical and thermomechanical phenomena occurring in the fuel rod during irradiation, it is necessary to perform calculations using computer codes. The ultimate goal is a description of fuel behaviour in both normal and abnormal conditions. From this knowledge, operating rules can be derived to prevent fuel failures and the release of fission products to the environment, and also, in extreme cases, to prevent escalation of fuel and core damage and the consequential hazards. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well-defined experiments suitable for code validation in association with the OECD/NEA; - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to give guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application

  16. The dependence of the global neutronic parameters on the fuel burnup for CANDU SEU43 core

    Energy Technology Data Exchange (ETDEWEB)

    Balaceanu, V. [Institute for Nuclear Research, Pitesti (Romania); Pavelescu, M. [Academy of Romanian Scientists, Bucharest (Romania)

    2010-05-15

    In order to reduce the total fuel costs for the CANDU reactors, mainly by extending the fuel burnup limits, some fuel bundle concepts have been developed in different CANDU owner countries. Therefore, in our Institute the SEU43 (Slightly Enriched Uranium with 43 fuel elements) project was started in early '90s. The neutronic behavior analysis of the CANDU core with SEU43 fuel was an important step in our project design. The objective of this paper is to highline an analysis of the neutronic behavior of the CANDU SEU43 core with the fuel burnup. More exactly, the study refers to the dependence of some global neutronic parameters, mainly the reactivity, on the fuel burnup. Two types of CANDU core were taken into consideration: reference core (without any reactivity devices) and perturbed core (with a strong reactivity system inserted). The considered reactivity system is the Mechanical Control Absorber (MCA) one. The performed parameters are: k{sub eff.} values, the MCA reactivity worth and flux distributions. The fuel bundles in the core are SEU43, with the fuel enrichment in U{sup 235} of 0.96% and at nominal power. For the fuel burnup the values are: 0.00 GWd/tU (fresh fuel); 8.00 GWd/tU and 25.00 GWd/tU. For reaching this objective, a global neutronic calculation system named WIMSPIJXYZ LEGENTR is used. Starting from a 69-groups ENDF/B-V based library, this system uses three transport codes: (1) the standard lattice-cell code WIMS, for generating macroscopic cross sections in supercell option and also for burnup calculations; (2) the PIJXYZ code for 3D simulation of the MCA reactivity devices and the 3D correction of the macroscopic cross sections; (3) the LEGENTR 3D transport code for estimating global neutronic parameters (CANDU core). The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON and DIREN codes. This comparison shows a good agreement between these results. (orig.)

  17. ELESTRES 2.1 computer code for high burnup CANDU fuel performance analysis

    International Nuclear Information System (INIS)

    The ELESTRES (ELEment Simulation and sTRESses) computer code models the thermal, mechanical and micro structural behaviours of CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains in fuel element design analysis and assessments. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. ELESTRES 2.1 was developed for high burnup fuel application, based on an industry standard tool version of the code, through the implementation or modification to code models such as fission gas release, fuel pellet densification, flux depression (radial power distribution in the fuel pellet), fuel pellet thermal conductivity, fuel sheath creep, fuel sheath yield strength, fuel sheath oxidation, two dimensional heat transfer between the fuel pellet and the fuel sheath; and an automatic finite element meshing capability to handle various fuel pellet shapes. The ELESTRES 2.1 code design and development was planned, implemented, verified, validated, and documented in accordance with the AECL software quality assurance program, which meets the requirements of the Canadian Standards Association standard for software quality assurance CSA N286.7-99. This paper presents an overview of the ELESTRES 2.1 code with descriptions of the code's theoretical background, solution methodologies, application range, input data, and interface with other analytical tools. Code verification and validation results, which are also discussed in the paper, have confirmed that ELESTRES 2.1 is capable of modelling important fuel phenomena and the code can be used in the design assessment and the verification of high burnup fuels. (author)

  18. A validated methodology for evaluating burnup credit in spent fuel casks

    International Nuclear Information System (INIS)

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

  19. Upgrading spent fuel shipping casks to meet higher burn-up

    International Nuclear Information System (INIS)

    In order to allow the transportability of high burn-up fuel and of MOX fuel in existing casks, TRANSNUCLEAIRE presents a two-step proven solution: (1) starting from 35/40 GWd/tU and 3.5 % enrichment, casks of the TN 12 family can be upgraded to 40/45 GWd/tU and 4.3 % enrichment by the use high performance baskets. (2) a second step consists in adding neutron shielding to allow transportation of fuel with a burn-up of 45/50 GWd/tU with a standard basket and of 50/55 GWd/tU with a high performance basket. (J.P.N.)

  20. On stability of spatial distributions of crystal structure defects in irradiated high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO2 fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO2 fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets

  1. A validated methodology for evaluating burnup credit in spent fuel casks

    International Nuclear Information System (INIS)

    The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in keff. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs

  2. Operational Experience of Nuclear Fuel in Finnish Nuclear Power Plants (with Emphasis on WWER Fuel)

    International Nuclear Information System (INIS)

    The four operating nuclear reactors in Finland, Loviisa-1 and -2 and Olkiluoto-1 and -2 have now operated approximately 30 years. The overall operational experience has been excellent. Load factors of all units have been for years among the highest in the world. The development of the fuel designs during the years has enabled remarkable improvement in the fuel performance in terms of burnup. Average discharge burnup has increased more than 30 percent in all Finnish reactor units. A systematic inspection of spent fuel assemblies, and especially all failed fuel assemblies, is a good and useful practise employed in Finland. A possibility to inspect the fuel on site using a pool side inspection facility is a relatively economic way to find out root causes of fuel failures and thereby facilitate developing remedies to prevent similar failures in the future

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

  4. Evaluation of the characteristics of high burnup and high plutonium content mixed oxide (MOX) fuel

    International Nuclear Information System (INIS)

    Two kinds of MOX fuel irradiation tests, i.e., MOX irradiation test up to high burnup and MOX having high plutonium content irradiation test, have been performed from JFY 2007 for five years in order to establish technical data concerning MOX fuel behavior during irradiation, which shall be needed in safety regulation of MOX fuel with high reliability. The high burnup MOX irradiation test consists of irradiation extension and post irradiation examination (PIE). The activities done in JFY 2011 are destructive post irradiation examination (D-PIE) such as EPMA and SIMS at CEA (Commissariat a l'Enegie Atomique) facility. Cadarache and PIE data analysis. In the frame of irradiation test of high plutonium content MOX fuel programme, MOX fuel rods with about 14wt % Pu content are being irradiated at BR-2 reactor and corresponding PIE is also being done at PIE facility (SCK/CEN: Studiecentrum voor Kernenergie/Centre d'Etude l'Energie Nucleaire) in Belgium. The activities done in JFY 2011 are non-destructive post irradiation examination (ND-PIE) and D-PIE and PIE data analysis. In this report the results of EPMA and SIMS with high burnup irradiation test and the result of gamma spectrometry measurement which can give FP gas release rate are reported. (author)

  5. Advances in applications of burnup credit to enhance spent fuel transportation, storage, reprocessing and disposition. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    Given a trend towards higher burnup power reactor fuel, the IAEA began an active programme in burnup credit (BUC) with major meetings in 1997 (IAEA-TECDOC-1013), 2000 (IAEA-TECDOC-1241) and 2002 (IAEA-TECDOC-1378) exploring worldwide interest in using BUC in spent fuel management systems. This publication contains the proceedings of the IAEA's 4th major BUC meeting, held in London. Sixty participants from 18 countries addressed calculation methodology, validation and criticality, safety criteria, procedural compliance with safety criteria, benefits of BUC applications, and regulatory aspects in BUC. This meeting encouraged the IAEA to continue its activities on burnup credit including dissemination of related information, given the number of Member States having to deal with increased spent fuel quantities and extended durations. A 5th major meeting on burnup credit is planned 2008. Burnup credit is a concept that takes credit for the reduced reactivity of fuel discharged from the reactor to improve loading density of irradiated fuel assemblies in storage, transportation, and disposal applications, relative to the assumption of fresh fuel nuclide inventories in loading calculations. This report has described a general four phase approach to be considered in burnup credit implementation. Much if not all of the background research and data acquisition necessary for successful burnup credit development in preparation for licensing has been completed. Many fuel types, facilities, and analysis methods are encompassed in the public knowledge base, such that in many cases this guidance will provide a means for rapid development of a burnup credit program. For newer assembly designs, higher enrichment fuels, and more extensive nuclide credit, additional research and development may be necessary, but even this work can build on the foundation that has been established to date. Those, it is hoped that this report will serve as a starting point with sufficient reference to

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

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M. (Argonne National Lab., IL (USA))

    1989-09-01

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

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

    International Nuclear Information System (INIS)

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

  8. Investigation of Irradiation Behavior of SiC-Coated Fuel Particle at Extended Burnup

    International Nuclear Information System (INIS)

    In current high-temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. The maximum burnup of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is limited to 3.6%FIMA (% fission per initial metallic atom) to certify its integrity during the operation. In order to investigate fuel behavior under extended burnup condition, irradiation tests were performed. The irradiation was carried out as HRB-22 and 91F-1A capsule irradiation tests. The fuel for the irradiation tests was called extended burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the HTTR. In order to keep fuel integrity up to over 5%FIMA, the thickness of buffer and SiC layers of fuel particle were increased. The fuel compacts were irradiated in the HRB-22 and the 91F-1A capsules at the High Flux Isotope Reactor of Oak Ridge National Laboratory and at the Japan Materials Testing Reactor of the Japan Atomic Energy Research Institute, respectively. The comparison of measured and calculated release rate-to-birth rate ratios showed that there were additional failures in both irradiation tests. A pressure vessel failure model analysis showed that no tensile stresses acted on the SiC layers even at the end of irradiation and no pressure vessel failure occurred in the intact particles even in a particle with thin buffer layer with failed OPyC layer. The presumed failure mechanisms are additional through-coatings failure of as-fabricated SiC-failed particles or an excessive increase of internal pressure

  9. Uranium and plutonium determinations for evaluation of high burnup fuel performance

    International Nuclear Information System (INIS)

    Purpose of this work is to experimentally test computational methods being developed for reactor fuel operation. Described are the analytical techniques used in the determination of uranium and plutonium compositions on PWR fuel that has spanned five power cycles, culminating in 55,000 to 57,000 MWd/T burnup. Analyses have been performed on ten samples excised from selected sections of the fuel rods. Hot cell operations required the separation of fuel from cladding and the comminution of the fuel. These tasks were successfully accomplished using a SpectroMil, a ball pestle impact grinding and blending instrument manufactured by Chemplex Industries, Inc., Eastchester, New York. The fuel was dissolved using strong mineral acids and bomb dissolution techniques. Separation of the fuel from fission products was done by solvent (hexone) extraction. Fuel isotopic compositions and assays were determined by the mass spectrometric isotope dilution (MSID) method using NBS standards SRM-993 and SRM-996. Alpha spectrometry was used to determine the 238Pu composition. Relative correlations of composition with burnup were obtained by gamma-ray spectrometry of selected fission products in the dissolved fuel

  10. Measurement techniques for verifying burnup

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.I. (Sandia National Lab., Albuquerque, NM (US)); Bierman, S.R. (Pacific Northwest Lab., Richland, WA (US))

    1992-05-01

    Measurements of the nuclear radiation from spent reactor fuel are being considered to qualify assemblies for loading into casks that will be used to transport spent fuel from utility sites to a federal storage facility. To ensure nuclear criticality safety, the casks are being designed to accept assemblies that meet restrictions as to burnup, initial enrichment and cooling time. This paper reports that measurements could be used to ensure that only fuel assemblies that meet the restrictions are selected for loading.

  11. Measurement techniques for verifying burnup

    International Nuclear Information System (INIS)

    Measurements of the nuclear radiation from spent reactor fuel are being considered to qualify assemblies for loading into casks that will be used to transport spent fuel from utility sites to a federal storage facility. To ensure nuclear criticality safety, the casks are being designed to accept assemblies that meet restrictions as to burnup, initial enrichment and cooling time. This paper reports that measurements could be used to ensure that only fuel assemblies that meet the restrictions are selected for loading

  12. A technique of melting temperature measurement and its application for irradiated high-burnup MOX fuels

    International Nuclear Information System (INIS)

    A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. (author)

  13. Characterisation of high-burnup LWR fuel rods through gamma tomography

    International Nuclear Information System (INIS)

    Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235U enrichment of the fresh UO2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

  14. Use of burnup credit in criticality evaluation for spent fuel storage pool

    International Nuclear Information System (INIS)

    Boraflex is a polymer based material which is used as matrix to contain a neutron absorber material, boron carbide. In a typical spent fuel pool the irradiated Boraflex has been known as a significant source of silica. Since 1996, it was reported that elevated silica levels were measured in the Ulchin Unit 2 spent fuel pool water. Therefore, the Ulchin Unit 2 spent fuel storage racks were needed to be reanalyzed to allow storage of fuel assemblies with normal enrichments up to 5.0w/o U-235 in all storage cell locations using credit for burnup. The analysis does not take any credit for the presence of the spent fuel rack Boraflex neutron absorber panels. In region 2, the calculations were performed by assuming in an infinite radial array of storage cells. No credit is taken for axial or radial neutron leakage. The water in the spent fuel storage pool was assumed to be pure. In the evaluation of the Ulchin Unit 2 spent fuel storage pool, criticality analyses were performed with the CASMO-3 code. A reactivity uncertainty in the fuel depletion calculations was combined with other calculational uncertainty. The manufacturing tolerances were considered, as well. From the calculation, the acceptable burnup domain in region 2 of the spent fuel storage pool. where the curve identifies conditions of equal reactivity for various initial enrichments between 1.6w/o and 5.0w/o, was evaluated. In region 2, the maximum keff including all uncertainties, is 0.94648 for the enrichment-burnup combination from loading curve. (author)

  15. Probabilistic safety criteria on high burnup HWR fuels

    International Nuclear Information System (INIS)

    BACO is a code for the simulation of the thermo-mechanical and fission gas behaviour of a cylindrical fuel rod under operation conditions. Their input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA-CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. The techniques for sensitivity analysis defined in BACO are: the 'extreme case analysis', the 'parametric analysis' and the 'probabilistic (or statistics) analysis'. We analyse the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs taking account probabilistic safety criteria and using the BACO's sensitivity analysis. (author)

  16. Assessment of dry storage performance of spent LWR fuel assemblies with increasing burnup

    International Nuclear Information System (INIS)

    To assess the extended storage performance of spent LWR-fuel, the available experience can be collated into 3 storage modes: mode I: fast decrease rate of temperature between maximum of licensed dry storage temperature and 300 deg. C; mode II: medium decrease rate of the fuel rod dry storage temperature between 300 deg. C and 200 deg. C; mode III: slow to negligible decrease rate of fuel rod dry storage temperature for temperatures less than 200 deg. C. Mode I is typical for early interim storage, mode III covers extremely long term storage which is encountered presumably for nearly all dry storage extensions to be considered. Mode II dry storage is characterised by the fact that all creep deformations of the spent fuel cladding can already be regarded as terminated as well as the corrosive attack of the cladding. Reviewing the fission product behaviour under dry storage conditions it can be pointed out that the fission products generated in the UO2-fuel under in service conditions are practically immobile in the UO2-fuel lattice during storage. Consequently all fission product driven defect mechanisms like stress corrosion cracking (SCC), uniform fuel rod internal fission product corrosion of the cladding, localised fuel rod internal fission product corrosion of the cladding, will not take place. The leading defect mechanism for spent fuel rod in dry storage - also for fuel rod with increased burn-up - remains creep due to the hoop strain resulting from the fuel rod internal fission gas pressure. Limiting the creep to its primary and secondary stages prevents fuel rod degradation. Post-pile creep of fuel rod cladding can be described conservatively by the creep of unirradiated cladding. The allowable uniform strain of the cladding in its typical post-pile condition preventing tertiary creep under dry spent fuel storage conditions is 1-2%. Dry storage performance prediction of fuel assemblies with a burn-up ≤ 65 GWd/tHM was calculated based on the fuel assemblies

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

    International Nuclear Information System (INIS)

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

  18. About the transportation of WWER-type reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    An intensive transition of existing NPPs with WWER-type reactors to new types of nuclear fuel and, therefore, to new fuel cycles, can be observed. New fuel cycles increase the burnup of nuclear fuel, and it increases the cost-effectiveness of its utilization

  19. Nuclear fuel element

    International Nuclear Information System (INIS)

    Purpose: To reduce the probability of stress corrosion cracks in a zirconium alloy fuel can even when tensile stresses are resulted to the fuel can. Constitution: Sintered nuclear fuel pellets composed of uranium dioxide or a solid solution of gadolinium as a burnable poison in uranium dioxide are charged in a tightly sealed zirconium alloy fuel can. The nuclear fuel pellets for the nuclear fuel element are heat-treated in a gas mixture of carbon dioxide and carbon monoxide. Further, a charging gas containing a mixture of carbon dioxide and carbon monoxide is charged within a zirconium alloy fuel can packed with the nuclear fuel pellets and tightly sealed. (Aizawa, K.)

  20. Analytical and numerical study of radiation effect up to high burnup in power reactor fuels

    International Nuclear Information System (INIS)

    In the present work the behavior of fuel pellets for power reactors in the high burnup range (average burnup higher than 50 MWd/kgHM) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup, as long as a new microstructure develops, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behaviour. The evolution of porosity in the high burnup structure (HBS) is assumed to be determinant of the retention capacity of the fission gases released by the matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Starting from several works published in the open literature, a model was developed to describe the behaviour and evolution of porosity at local burnup values ranging from 60 to 300 MWd/KgHM. The model is mathematically expressed by a system of non-linear differential equations that take into account the open and closed porosity, the interactions between pores and the free surface and phenomena like pore's coalescence and migration and gas venting. Interactions of different orders between open and closed pores, growth of pores radius by vacancies trapping, the evolution of the pores number density, the internal pressure and over pressure within the pores, the fission gas retained in the matrix and released to the free volume are analyzed. The results of the simulations performed in the present work are in excellent agreement with experimental data available in the open literature and with results calculated by other authors (author)

  1. 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.

  2. Flowchart evaluations of irradiated fuel treatment process of low burnup thorium

    International Nuclear Information System (INIS)

    A literature survey has been carried out, on some versions of the acid-thorex process. Flowsheets of the different parts of the process were evaluated with mixer-settlers experiments. A low burnup thorium fuel (mass ratio Th/U∼100/1), proposed for Brazilian fast breeder reactor initial program, was considered. The behaviour of some fission products was studied by irradiated tracers techniques. Modifications in some of the process parameters were necessary to achieve low losses of 233U and 232U and 232Th. A modified acid-thorex process flowsheet, evaluated in a complete operational cycle, for the treatment of low burnup thorium fuels, is presented. High decontamination factors of thorium in uranium, with reasonable decontamination of uranium in thorium, were achieved. (author)

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

    International Nuclear Information System (INIS)

    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 UO2 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)

  4. Determination of dependence of fissile fraction in MOX fuels on spent fuel storage period for different burnup values

    International Nuclear Information System (INIS)

    Highlights: ► In a previous study, an expression to calculate fissile fraction of MOX for various burnups was obtained for 5-year cooled SF. ► In this follow-up study, a correction factor for spent fuel storage periods other than 5 years is derived. ► Thus, one major restriction on use of the expression derived in the initial study is eliminated. - Abstract: The purpose of this technical note is to remove one of the limitations of a derived expression in a previously published article (Özdemir et al., 2011). The original article focused on deriving (computationally) an expression for calculating total fissile fraction of mixed oxid (MOX) fuels depending on discharge burnup of spent fuel and desired burnup of MOX fuel; consequently, such an expression was obtained and put forward, together with its limitations. One of the limitations has been that all the computations and therefore the resulting expression are based on the assumption of a spent fuel storage period of 5 years. This follow-up study simply aims to obtain a correction factor for spent fuel storage periods other than 5 years; thus to remove one major restriction on use of the expression derived in the original article

  5. Estimate of fuel burnup spatial a multipurpose reactor in computer simulation; Estimativa da queima espacial do combustivel de um reator multiproposito por simulacao computacional

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

  6. Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks

    International Nuclear Information System (INIS)

    This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs

  7. Material requirements for a thorium based nuclear fuel

    OpenAIRE

    Galiana Gonzalez, Bernat

    2010-01-01

    The increase in the energy consumption and the expected growth in the nuclear capacity make it necessary to look for alternative fuels to replace uranium. The fuel chosen, which was also considered in the early stages of nuclear energy, is thorium. Thorium has some characteristics that make it valuable as a fuel, like its abundance, the low radiotoxicity of the waste generated, the higher economy regarding its larger absorption cross-section and higher burnups and the prolifera...

  8. Optimization of the Trasuranus Burnup Model for GD-Doped WWER-1000 Fuel Pins Based on Results of HELIOS Code

    International Nuclear Information System (INIS)

    This paper describes results of testing of the TRANSURANUS burn-up model (TUBRNP routine) for Gd-doped WWER-1000 fuel pin based on results of HELIOS code. The testing covers the analysis of different types of nuclear fuel rods from a neutronic point of view that one can encounter in the VVER-1000 reactor core. The HELIOS computations simulate the assembly geometry, and combine 4 different 235U enrichment configurations with 4 different Gd2O3-concentrations. For each of these combinations the radial distribution of the concentrations of 155Gd and 157Gd compute in one Gd-doped rod. Based on these results the recommendations on using cross section of Gd in TRANSURANUS TUBRNP model were proposed. (author)

  9. Grain size and burnup dependence of spent fuel oxidation: Geological repository impact

    International Nuclear Information System (INIS)

    Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate and an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent-fuel samples oxidized in thermogravimetric analysis (TGA) or oven dry-bath (ODB) experiments. The experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U4O9r↓U3O8. Model results predict that U3O8 formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficiently low

  10. Modeling fission gas release in high burnup ThO2-UO2 fuel

    International Nuclear Information System (INIS)

    A preliminary fission gas release model to predict the performance of thoria fuel using the FRAPCON-3 computer code package has been formulated. The following modeling changes have been made in the code: - Radial power/burnup distribution; - Thermal conductivity and thermal expansion; - Rim porosity and fuel density; - Diffusion coefficient of fission gas in ThO2-UO2 fuel and low temperature fission gas release model. Due to its lower epithermal resonance absorption, thoria fuel experiences a much flatter distribution of radial fissile products and radial power distribution during operation as compared to uranian fuel. The rim effect and its consequences in thoria fuel, therefore, are expected to occur only at relatively high burnup levels. The enhanced conductivity is evident for ThO2, but for a mixture the thermal conductivity enhancement is small. The lower thermal fuel expansion tends to negate these small advantages. With the modifications above, the new version of FRAPCON-3 matched the measured fission gas release data reasonably well using the ANS 5.4 fission gas release model. (authors)

  11. Chemical separation for the burnup determination of the U3Si/Al spent fuels

    International Nuclear Information System (INIS)

    The separation of U, Pu, and Nd for the burnup determination of the U3Si/Al spent fuel samples has been studied. The preliminary experiments were carried out with the simulated spent fuel solution. The solutions were prepared by adding of fission product elements to unirradiated U3Si/Al fuel samples. The fuel samples were dissolved in 6 M HNO3, 6 M HNO3 using mercury catalyst, or applying a mixture of HCl and HNO3 without any catalyst. All dissolved fuel solutions contained a small amount of a residue(silica). The trace silica reprecipitated from the fuel solutions taken for the separation was dissolved in HF and removed by subsequent evaporating to dryness. The separation of U and fission product elements from the various sample solutions was achieved by two sequential anion exchange resin separation procedures. The U, Pu and Nd can be purely isolated from the sample solutions with a large excess of Al by this chromatographic procedures. The dissolution and separation procedure used in this experiment were applied for burnup determination of real U3Si/Al spent fuels from HANARO reactor

  12. Chemical analyses and calculation of isotopic compositions of high-burnup UO2 fuels and MOX fuels

    International Nuclear Information System (INIS)

    Chemical analysis activities of isotopic compositions of high-burnup UO2 fuels and MOX fuels in CRIEPI and calculation evaluation are reviewed briefly. C/E values of ORIGEN2, in which original libraries and JENDL-3.2 libraries are used, and other codes with chemical analysis data are reviewed and evaluated. Isotopic compositions of main U and Pu in fuels can be evaluated within 10% relative errors by suitable libraries and codes. Void ratio is effective parameter for C/E values in BWR fuels. JENDL-3.2 library shows remarkable improvement compared with original libraries in isotopic composition evaluations of FP nuclides. (author)

  13. Development of a code and models for high burnup fuel performance analysis

    International Nuclear Information System (INIS)

    First the high burnup LWR fuel behavior is discussed and necessary models for the analysis are reviewed. These aspects of behavior are the changes of power history due to the higher enrichment, the temperature feedback due to fission gas release and resultant degradation of gap conductance, axial fission gas transport in fuel free volume, fuel conductivity degradation due to fission product solution and modification of fuel micro-structure. Models developed for these phenomena, modifications in the code, and the benchmark results mainly based on Risoe fission gas project is presented. Finally the rim effect which is observe only around the fuel periphery will be discussed focusing into the fuel conductivity degradation and swelling due to the porosity development. (author). 18 refs, 13 figs, 3 tabs

  14. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this

  15. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    International Nuclear Information System (INIS)

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from ∼ ±40% at beginning of life to ∼ ±10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this case, a self

  16. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  17. The formation process of the pellet-cladding bonding layer in high burnup BWR fuels

    International Nuclear Information System (INIS)

    The bonding formation process was studied by EPMA analysis, XRD measurements, and SEM/TEM observations for the oxide layer on a cladding inner surface and the pellet-cladding bonding layer in irradiated fuel rods. Specimens were prepared from fuels which had been irradiated to the pellet average burnups of 15, 27, 42 and 49 GWd/t in BWRs. In the lower burnup specimens of 15 and 27 GWd/t, no bonding layer was found, while the higher burnup specimens of 42 and 49 GWd/t had a typical bonding layer about 10 to 20 μm thick. A bonding layer which consisted of two regions was found in the latter fuels. One region of the inner surface of the Zr liner cladding was made up mainly of ZrO2 with a small amount of dissolved UO2. The structure of this ZrO2 consisted of cubic polycrystals a few nanometers in size, while no monoclinic crystals were found. The other region, near the pellet surface, had both a cubic solid solution of (U,Zr)O2 and amorphous phase in which the concentrations of UO2 and ZrO2 changed continuously. Even in the lower burnup specimens having no bonding layer, cubic ZrO2 phase was identified in the cladding inner oxide layer. The XRD measurements were consistent with the TEM results of the absence of the monoclinic ZrO2 phase. Phase transformation and amorphization were attributed to fission damage, since such phenomena have never been observed in the cladding outer surface. Phase transformation from monoclinic to cubic ZrO2 and amorphization by irradiation damage of fission products were discussed in connection with the formation mechanism and conditions of the bonding layer. (author)

  18. Burnup determination of power reactor fuel elements by gamma spectrometry

    International Nuclear Information System (INIS)

    This report describes a method for determining by γ spectrometry the burn up and the specific power of fuel elements irradiated in power reactors. The energy spectrum of γ rays emitted by fission products is measured by means of a simple equipment using a sodium iodide detector and a multichannel analyzer. In order to extract from the spectrum a quantity proportional to the burn up, it is necessary to: - isolate an activity specific of one emitter,- give the same importance to fissions in uranium and plutonium - take into account the radioactive decay during and after irradiation. One hundred fuel elements were studied and burn up values obtained by γ spectrometry are compared to results given by chemical analyses. Preliminary measurements show that the accuracy of the results is greatly increased by the use of a germanium detector, due to its good resolution. (authors)

  19. Validation of SWAT for burnup credit problems by analysis of post irradiation examination of 17*17 PWR fuel assembly

    International Nuclear Information System (INIS)

    For adopting burnup credit in transport or storage of spent fuel (SF), development of a reliable burnup calculation code is crucial. For this purpose, data of Post Irradiation Examination (PIE) have been extensively analyzed to evaluate accuracy of burnup calculation codes for a 14*14 or 15*15 PWR fuel assembly. This study shows results of analysis of this latest PIE with SWAT and ORIGEN2.1. SWAT is an integrated burnup code system for a 17*17 PWR fuel assembly that has been developed by Tohoku University and JAERI. The results show that SWAT can more precisely predict nuclide composition of latest PWR assembly than ORIGEN2.1. (O.M.)

  20. Nuclear fuel transporting container

    International Nuclear Information System (INIS)

    Purpose: To prevent the failure of nuclear fuel rods constituting a nuclear fuel assembly contained to the inside of a container upon fire accidents or the likes. Constitution: The nuclear fuel transportation container comprises a tightly sealed inner vessel made of steels for containing a nuclear fuel assembly consisting of bundled nuclear fuel rods, a heat shielding material surrounding the inner vessel, shock absorber and an outer vessel. A relief safety valve is disposed to the inner vessel that actuates at a specific pressure higher than the normal inner pressure for the nuclear fuel rods of the fuel assembly and lower than the allowable inner pressure of the inner vessel. The inside of the inner vessel is pressurized by way of the safety valve such that the normal inner pressure in the inner vessel is substantially equal to the normal inner pressure for the nuclear fuel rods. (Aizawa, K.)

  1. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

    2014-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  2. Simulation of fuel cycles with minor actinide management using a fast burnup calculation tool

    International Nuclear Information System (INIS)

    The paper presents a fast and flexible burnup model for fuel cycle simulations which is based on the description of the one-group cross-sections as analytic functions of the isotopic composition. This was accomplished by multi-dimensional regression based on the results of numerous core calculations. The developed model is able to determine the spent fuel composition in reasonable CPU time, and was integrated into a simplified fuel cycle model containing Gas Cooled Fast Reactors (GFR) and conventional light water reactors (LWRs). The fuel cycle simulations revealed an advantageous effect of increased minor actinide content in the GFR core on the fuel utilization parameters. In order to explore the processes that lay behind this effect the neutronics balance of the GFR was investigated in equilibrium cycle conditions. (author)

  3. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  4. PWR AXIAL BURNUP PROFILE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  5. Extended Burnup Impact on the TN24 Spent Fuel Storage Cask Main Parameters

    International Nuclear Information System (INIS)

    In order to establish the capability of the TN24 cask for storage of spent fuel assemblies which are beyond the limits given by the manufacturer, a calculations of the dose and heat decay have been made for several cases of burnup higher than 35 GWd/MTU, using the SCALE 4.2 code package. The results were compared with the data obtained from the manufacturer. According to the results of the ORIGEN and SAS4 calculations and taking into the account limitations of the used model, it is possible to estimate that for 50 GWd/MTU burnup at least 15 years cooling time period is necessary to allow the use of TN24 cask. (author)

  6. ZADOC, 2 Group Time-Dependent Burnup in X-Y Geometry with Fuel Management

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: Two neutron group diffusion equations for a square mesh in x-y geometry are solved to yield a power distribution. Burnup for one time step is simulated by interpolation in a library of two-group cross sections which forms part of the problem data. It is assumed that the power distribution is invariant during one time step, at the end of which a re-calculation of flux and power follows automatically. Burnup proceeds in a succession of time steps and a number of fuel management options are available. 2 - Method of solution: Standard finite difference methods are used. 3 - Restrictions on the complexity of the problem: The programme is restricted to two neutron groups. Limits on the number of mesh points are as follows - IBM 7030 61 x 61; IBM 7090 32 x 32; see also AEEW - R.425

  7. IFPE/AECL-BUNDLE, Fission Gas Release and Burnup Analysis, PHWR Fuel

    International Nuclear Information System (INIS)

    Description: Prototype Candu Fuel bundles for the CANDU6 (bundle NR) and Bruce (bundle JC) reactors were irradiated in the NRU experimental reactor at Chalk River Laboratories in experimental loop facilities under typical Candu reactor conditions, except that they were cooled using light water. NEA-1596/01 - Description: Bundle JC was a prototype 37-element fuel bundle for the Bruce-A Ontario Hydro reactors. This pressurized heavy water reactor (PHWR) design utilizes a heavy water moderator and pressurize heavy water coolant. For irradiation in the NRU reactor, the centre fuel element was removed and replaced by a central tie rod for irradiation purposes in the vertical test section. Coolant for the test was pressurized light water under typical PHWR conditions of 9 to 10.5 MPa and 300 deg. C. The fuel elements used 1.55 wt% U-235 in U uranium dioxide fuel and were clad with Zircaloy-4 material. The bundles' elements were coated with a graphite coating. The fuel is somewhat atypical of 37 element-type fuel since the length to diameter ratio (l/d) is large (1.73) due to the pellets being ground down from a OD of 14.3 mm to 12.12 mm. The outer element burnup averaged approximately 640 MWh/kgU on discharge. Outer element powers varied between 57 kW/m near the beginning of life and 23 kW/m at discharge. Due to the long irradiation, the bundle experienced 153 short shutdowns, and 129 longer duration shutdowns. No element instrumentation was used during the irradiation. However, the bundle was subjected to extensive post-irradiation examination (PIE) that included dimensional changes, fission gas release, fuel burnup analysis, and metallography that included grain size measurement. NEA-1596/02 - Description: Bundle NR was a prototype 37-element fuel bundle for the Candu 600 reactor. This pressurized heavy water reactor (PHWR) design utilizes a heavy water moderator and pressurized heavy water coolant. For irradiation in the NRU reactor, the centre fuel element was

  8. World nuclear fuel cycle requirements 1990

    International Nuclear Information System (INIS)

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management

  9. Nuclear fuel cycles of WWER-1000 at Kozloduy NPP: a program for transition to advanced fuel

    International Nuclear Information System (INIS)

    A systematical approach for WWER nuclear fuel utilization improvement is applied at the Kozloduy NPP while observing safety requirements. The fuel assemblies utilization for 4 years, higher burnup and reducing of high-activity radioactive waste will be achieved by realization of Activities program for transition of Units 5 and 6 of Kozloduy NPP to operating with Alternative Fuel Assemblies (AFA). The specified results from advanced fuel introduction also will improve the economic indices of Units 5 and 6 operation at Kozloduy NPP

  10. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    International Nuclear Information System (INIS)

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring

  11. FUMEX-III: A New IAEA Coordinated Research Project on Fuel Modelling at Extended Burnup

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency has initiated a new a Coordinated Research Project on Fuel Modelling at Extended Burnup (FUMEX-III). Currently, thirty one fuel modelling groups are participating with the intention of improving their capabilities to understand and predict the behaviour of water reactor fuel at high burnups. The exercise is carried in coordination with the OECD/NEA. The participants will model test cases provided by from sources such as the Halden Reactor Project and commercial irradiations and tests from the participants themselves. It is also intended to utilise idealised cases to test model behaviour under high burnup conditions. All cases are maintained in the OECD International Fuel Performance Experimental (IFPE) Database. The participants are particularly interested in modelling transient behaviour and mechanical interactions between pellet and cladding, including severe transient behaviour (RIA/LOCA) as well as temperature and fission gas release. However the participants include newcomer teams as well as state-of-the-art code users and have differing needs depending on the reactor system that they are modelling (PHWR, PWR, BWR, WWER) and the level of code development and experience that they have, so a matrix of test cases has been developed to allow each team to test their codes and methods appropriately. Some codes (eg TRANSURANUS and FEMAXI) are being used by several teams, both developing models and code user expertise. This paper summarises the objectives of the participants, the matrix of test cases that has been made available to the participants and some additional cases that are being prepared for inclusion in the later stages of the Project. (authors)

  12. Results of post-irradiation examination to validate WWER-440 and WWER-1000 fuel efficiency at high burnups

    International Nuclear Information System (INIS)

    During the last 10 years on the basis of commercial operation of WWER reactors, a conversion from three to four year fuel cycle operation has been succeeded for WWER-440 fuel. This paper presents the examinations of fuel rods and fuel assemblies operated at different NPPs of Russia and Eastern Europe. Three WWER-440 fuel assemblies with different burnups and different irradiation in the core (3, 4 and 5 years fuel cycle) have passed full-scale examinations including both: destructive and non-destructive methods. The results of examinations have revealed that the irregularity of the field of the energy release may result in increased cladding oxidation. A validation of WWER-440 and WWER-1000 fuel efficiency during 4 and 5 years fuel cycles is also made on the basis of assessment of fuel rod and assembly mechanical state and changes in their geometry. The status of the fuel column including grain size, fuel swelling, rim layer and fission gas release depending on fuel burnup are investigated. During examinations mechanical properties and oxidation of the cladding, mechanical and corrosion state of the spacer grid, ultimate stress, hardness and plasticity of central tube and relaxation of spring unit are studied at the maximal fuel burnup 64 MWd/kgU for WWER-440 and 58 MWd/kgU for WWER-1000. Based on the examination results for the principal parameters determined fuel resource (variation in form, material structure and properties, corrosion resistance of the claddings, FGR form the fuel, fuel cladding interaction degree) reliable fuel operation at the burnups corresponding to four and five fuel cycles may be predicted. None of fuel efficiency factors in up-to date FA design are limited for operation during five fuel cycles

  13. A method for VVER-1000 fuel rearrangement optimization taking into account both fuel cladding durability and burnup

    Energy Technology Data Exchange (ETDEWEB)

    Pelykh, S.N., E-mail: 1@pelykh.net [Odessa National Polytechnic University, Shevchenko av., 1, Odessa 65044 (Ukraine); Maksimov, M.V. [Odessa National Polytechnic University, Shevchenko av., 1, Odessa 65044 (Ukraine); Parks, G.T. [University of Cambridge, Engineering Department, Trumpington str., Cambridge, CB2 1PZ (United Kingdom)

    2013-04-15

    Highlights: ► Fuel cladding life and burnup determine fuel rearrangement optimization. ► The proposed rearrangement optimization criterion is universal. ► Robust optimization conservatively confirms the outcome of deterministic optimization. ► Robust optimization reveals the trade-off between mean performance and consistency of performance in the face of uncertainties in operational conditions. -- Abstract: A method for VVER-1000 fuel rearrangement optimization that takes into account both cladding durability and fuel burnup and which is suitable for any regime of normal reactor operation has been established. The main stages involved in solving the problem of fuel rearrangement optimization are discussed in detail. Using the proposed fuel rearrangement efficiency criterion, a simple example VVER-1000 fuel rearrangement optimization problem is solved under deterministic and uncertain conditions. It is shown that the deterministic and robust (in the face of uncertainty) solutions of the rearrangement optimization problem are similar in principle, but the robust solution is, as might be anticipated, more conservative.

  14. A method for VVER-1000 fuel rearrangement optimization taking into account both fuel cladding durability and burnup

    International Nuclear Information System (INIS)

    Highlights: ► Fuel cladding life and burnup determine fuel rearrangement optimization. ► The proposed rearrangement optimization criterion is universal. ► Robust optimization conservatively confirms the outcome of deterministic optimization. ► Robust optimization reveals the trade-off between mean performance and consistency of performance in the face of uncertainties in operational conditions. -- Abstract: A method for VVER-1000 fuel rearrangement optimization that takes into account both cladding durability and fuel burnup and which is suitable for any regime of normal reactor operation has been established. The main stages involved in solving the problem of fuel rearrangement optimization are discussed in detail. Using the proposed fuel rearrangement efficiency criterion, a simple example VVER-1000 fuel rearrangement optimization problem is solved under deterministic and uncertain conditions. It is shown that the deterministic and robust (in the face of uncertainty) solutions of the rearrangement optimization problem are similar in principle, but the robust solution is, as might be anticipated, more conservative

  15. Nuclear fuel cycles

    International Nuclear Information System (INIS)

    The source of energy in the nuclear reactors in fission if a heavy nuclei by absorbing a neutron and giving fission products, few neutrons and gamma radiation. The Nuclear Fuel Cycle may be broadly defined as the set of process and operations needed to manufacture nuclear fuels, to irradiate them in nuclear reactors and to treat and store them, temporarily or permanently, after irradiation. Several nuclear fuel cycles may be considered, depending on the type of reactor and the type of fuel used and whether or not the irradiated fuel will be reprocessed. The nuclear fuel cycle starts with uranium exploration and ends with final disposal of the material used and generated during the cycle. For practical reasons the process has been further subdivided into the front-end and the back-end. The front-end of the cycle occurs before irradiation and the back-end begins with the discharge of spent fuel from the reactor

  16. Nuclear fuel lease accounting

    International Nuclear Information System (INIS)

    The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

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

    International Nuclear Information System (INIS)

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

  18. Monte Carlo burnup analysis code development and application to an incore thermionic space nuclear power system

    International Nuclear Information System (INIS)

    In the design of the incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI), the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional burnup and fuel depletion analysis codes have been found to be inadequate for these calculations, largely because of the material and geometry modeled and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. More sophisticated codes such as the transport lattice type code WIMS often lack some materials, such as ZrH. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the fuel element is needed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations. This effort required the modification of MCNP itself to perform the additional task of accomplishing burnup calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up

  19. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    The traditional assumption used in evaluating criticality safety of spent fuel cask is that the spent fuel is as reactive as when it was fresh (new). This is known as the fresh fuel assumption. It avoids a number of calculational and verification difficulties, but could take a heavy toll in decreased efficiency. The alternative to the fresh fuel assumption is called burnup credit. That is, the reduced reactivity of spent fuel that comes about from depletion of fissile radionuclides and net increase in neutron absorbers (poisons) is taken into account. It is recognizable that the use of burnup credit will in fact increase the percentage of unacceptable or non-specification fuel available for misloading. This could reduce individual cask safety margins if current practices with respect to loading procedures are maintained. As such, additional operational, design, analysis, and validation requirements should be established that, as a minimum, compensate for any potential reduction in fuel loading safety margin. This method is based on a probabilistic (PRA) approach and is called a relative risk comparison. The method assumes a linear risk model, and uses a selected probability function to compare the system of interest and an acceptable reference system by varying the features of each to assess effects on system safety. While risk is the product of an event probability and its consequence, the consequences of criticality in a cask are considered to be both unacceptable and the same, regardless of the initiating sequence. Therefore, only the probability of the event is considered in a relative risk evaluation

  20. Investigation of the CANLUB/sheath interface in CANDU fuel at extended burnup by XPS and SEM/WDX

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Behnke, R.; Duclos, A.M.; Gerwing, A.F. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Chan, P.K. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    1997-07-01

    A systematic investigation of the fuel-sheath interface in CANDU fuel as a function of extended burnup has been undertaken by XPS and SEM/WDX analysis. Adherent deposits of UO{sub 2} and fission products, including Cs, Ba, Rb, I, Te, Cd and possibly Ru, have been routinely identified on CANLUB coated and bare Zircaloy surfaces. Some trends in the distribution and chemistry of key fission products have begun to emerge. Several potential mechanisms for degradation of the CANLUB graphite layer at high burnup have been practically excluded. New evidence of carbon relocation within the fuel element and limited reaction with excess oxygen has also been obtained. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-15

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

  2. Spent fuel management options and nuclear fuel supplies in Germany

    International Nuclear Information System (INIS)

    The spent fuel management pathway adopted has a direct bearing on the supply of nuclear fuel. Compared to direct disposal, reprocessing is able to reduce the consumption of uranium, thus making nuclear power a quasi-indigenous source of power. The breeder technology was developed to make use of as many fuel constituents of natural uranium as possible, especially Pu-239. When used in mixed oxide fuel assemblies, plutonium can be burnt even in light water reactors. On the basis of three different scenarios for the development of the installed nuclear generating capacity, the annual uranium requirement up to 2030 is simulated in a computer model. The parameters influencing the calculation are the time, final storage, reprocessing, the use of mixed oxide fuel, and a higher fuel burnup. The service life of a nuclear power plant is assumed to be 35 years throughout. All steps of the nuclear fuel cycle are modeled, from purchasing the natural uranium to final storage. In each of the three scenarios, the model calculations arrive at clearly lower prices of natural uranium, of approx. US Dollar 65/kg of U, than actually prevailed in the second half of the seventies, i.e. more than US Dollar 190/kg of U. (orig.)

  3. Burnup determination in irradiated fuel by means of isotopic analysis and comparison to CASMO calculations

    International Nuclear Information System (INIS)

    One of the traditional methods for determining the burnup of irradiated Light Water Reactor (LWR) fuel is the 148Nd method according to ASTM E-321. Probably one of the largest sources for systematic errors in this method is the assumed fission yield, requiring knowledge of the fraction of fissions occurring in different fissile nuclides. Another traditional method for burnup determination is based on the uranium and plutonium isotopic composition; however, this method is rarely used for LWR fuel due to its rather simplified and rough assumptions regarding the neutron spectrum and fission fractions. However, modern physics codes like CASMO and HELIOS are instead able to calculate the amount of fission products and actinides formed or consumed during reactor operation in a much more sophisticated way. Isotopic Dilution Analysis with chemical separation of elements of interest, followed by isotopic analysis with a Thermal Ionization Mass Spectrometer (TIMS) is a well established method for determining the content of selected isotopes in samples of dissolved irradiated fuel. This method normally provides very accurate and precise results. High Performance Liquid Chromatography (HPLC) for elemental separations, combined with Inductively Coupled Plasma Mass Spectrometry (ICP-MS) has become a much faster alternative. In general, this method is somewhat less precise. This disadvantage is at least partly compensated by the possibility of analyzing a larger number of nuclides and samples. The local pellet burnup of a well characterised fuel sample irradiated in the Swedish Boiling Water Reactor Forsmark 3 to about 60 MWd/kgU was determined. Weight ratios of neodymium isotopes relative to 238U, analysed by Isotope Dilution Analysis applying HPLC-ICP-MS as well as 235U and 239Pu abundance values were compared to corresponding values calculated by a single-assembly CASMO-4 simulation. Input data were generated by CASMO-4/POLCA7 core tracking calculations. The overall result

  4. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    OpenAIRE

    Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2015-01-01

    A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the n...

  5. Highlights on R and D work related to the achievement of high burnup with MOX fuel in commercial reactors

    International Nuclear Information System (INIS)

    Part of the R and D work made at BELGONUCLEAIRE in the field of high burnup achievement with MOX fuel in commercial LWRs is made through lnternational Programmes. Special attention is given to the evolution with burnup of fuel neutronic characteristics and of in-reactor rod thermal-mechanical behaviour. Pu burning in MOX is characterized essentially by a drop of Pu239 content. The other Pu isotopes have an almost unchanged concentration, due to internal breeding. The reactivity drop of MOX versus burnup is consequently much less pronounced than in UO2 fuel. Concentration of minor actinides Am and Cm becomes significant with burnup increase. These nuclides start to play a role on total reactivity and in the helium production. The thermal-mechanical behaviour of MOX fuel rod is very similar to that of UO2. Some specificities are noticed. The better PCI resistance recognized to MOX fuel has recently been confirmed. Three PWR MOX segments pm-irradiated up to 58 GWd/tM were ramped at 100 W/cm.min respectively to 430-450-500 W/cm followed by a hold time of 24 hours. No segment failed. MOX and UO2 fuels have different reactivities and operate thus at different powers. Moreover, radial distribution of power in MOX pellet is less depressed at high burnup than in UO2, leading to higher fuel central temperature for a same rating. The thermal conductivity of MOX fuel decreases with Pu content, typically 4% for 10% Pu. The combination of these three elements (power level, power profile, and conductivity) lead to larger FGR at high burnup compared to UO2. Helium production remains low compared to fission gas production (ratio < 0.2). As faster diffusing element, the helium fractional release is much higher than that of fission gas, leading to rod pressure increase comparable to the one resulting from fission gas. (author)

  6. World nuclear fuel cycle requirements 1989

    International Nuclear Information System (INIS)

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under two nuclear supply scenarios. These two scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries in the World Outside Centrally Planned Economic Areas (WOCA). A No New Orders scenarios is also presented for the Unites States. This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the WOCA projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel; discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2020 for the Lower and Upper Reference cases and through 2036, the last year in which spent fuel is discharged, for the No New Orders case

  7. Overview of the burnup credit activities of the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA)

    International Nuclear Information System (INIS)

    This article summarises activities of the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) Expert Group on Burnup Credit Criticality, a subordinate group to the Working Part on Nuclear Criticality Safety (WPNCS). The WPNCS of the OECD/NEA coordinates and carries out work in the domain of criticality safety at the international level. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burnup credit. The activities of the expert group are aimed toward improving safety and identifying economic solutions to issues concerning the back-end of the fuel cycle. The main objective of the activities of the OECD/NEA Expert Group on Burnup Credit Criticality is to demonstrate that the available criticality safety calculational tools are appropriate for application to irradiated (burned) nuclear fuel systems and that a reasonable safety margin can be established. The method established by the expert group for investigating the physics and predictability of burnup credit is based on the specification and comparison of calculational benchmark problems. A wide range of fuel types, including PWR, BWR, MOX, and VVER fuels, has been or is being addressed by the expert group. The objective and status of each of these benchmark problems is reviewed in this article. It is important to note that the focus of the expert group is the comparison of the results submitted by each participant to assess the capability of commonly used code systems, not to quantify the physical phenomena investigated in the comparisons or to make recommendations for licensing action. (author)

  8. New Fuel Alloys Seeking Optimal Solidus and Phase Behavior for High Burnup and TRU Burning

    International Nuclear Information System (INIS)

    Recent modifications to fast reactor metallic fuels have been directed toward improving the melting and phase behaviors of the fuel alloy, for the purpose of ultra-high burnup and transuranic (TRU) burning. Improved melting temperatures increase the safety margin for uranium-based fast reactor fuel alloys, which is especially important for transuranic burning because the introduction of plutonium and neptunium acts to lower the alloy melting temperature. Improved phase behavior—single-phase, body-centered cubic—is desired because the phase is isotropic and the alloy properties are more predictable. An optimal alloy with both improvements was therefore sought through a comprehensive literature survey and theoretical analyses, and the creation and testing of some alloys selected by the analyses. Summarized here are those analyses, the impact of alloy modifications, and recent experimental results for selected pseudo-binary alloy systems that are hoped to accomplish the goals in a short timeframe. (author)

  9. REBUS: A burnup credit experimental programme

    International Nuclear Information System (INIS)

    An international programme called REBUS (REactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK-CEN and Belgonucleaire. At present it is sponsored by USNRC, EdF from France and VGB, representing German nuclear utilities. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark would qualify the codes to perform calculations of the burn-up credit. The benchmark exercise will investigate the following fuel types with associated burn-up. 1. Reference absorber test bundle, 2. Fresh commercial PWR UO2 fuel, 3. Irradiated commercial PWR UO2 fuel (50 GWd/tM), 4. Fresh PWR UO2 fuel, 5. Irradiated PWR UO2 fuel (30 GWd/tM). Reactivity effects will be measured in the critical facility VENUS. The accumulated burn-up of all rods will be measured non-destructively by gamma-spectrometry. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-18 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). The experimental implementation of the programme will start in 2000. (author)

  10. Analysis of high burnup fuel behavior under rod ejection accident in the Westinghouse-designed 950 MWe PWR

    International Nuclear Information System (INIS)

    As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident (RIA) may occur at the energy lower than the expected, duel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod burnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the core is less than 4 percent. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied. (author)

  11. LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System

    International Nuclear Information System (INIS)

    1 - Description of program or function: The LOLA-SYSTEM is a part of the JEN-UPM code package for PWR fuel management, scope or design calculations. It is a code package for core burnup calculations using nodal theory based on a FLARE type code. The LOLA-SYSTEM includes four modules: the first one (MELON-3) generates the constants of the K-inf and M2 correlations to be input into SIMULA-3. It needs the K-inf and M2 fuel assembly values at different conditions of moderator temperature, Boron concentration, burnup, etc., which are provided by MARIA fuel assembly calculations. The main module (SIMULA-3) is the core burnup calculation code in three dimensions and one group of energy. It normally uses a geometrical representation of one node per fuel assembly or per quarter of fuel assembly. It has included a thermal hydraulic feedback on flow and voids and criticality searches on boron concentration and control rods insertion. The CONCON code makes the calculation of the albedo, transport factors, K-inf and M2 correction factors to be input into SIMULA-3. The calculation is made in the XY transversal plane. The CONAXI code is similar to CONCON, but in the axial direction. 2 - Method of solution: MELON-3 makes a mean squares fit of K-inf and M2 values at different conditions in order to determine the constants of the feedback correlations. SIMULA-3 uses a modified one-group nodal theory, with a new transport kernel that provides the same node interface leakages as a fine mesh diffusion calculation. CONCON and CONAXI determine the transport and correction factors, as well as the albedo, to be input into SIMULA-3. They are determined by a method of leakages equivalent to the detailed diffusion calculation of CARMEN or VENTURE; these factors also include the heterogeneity effects inside the node. 3 - Restrictions on the complexity of the problem: Number of axial nodes less than or equal 34. Number of material types less than or equal 30. Number of fuel assembly types less

  12. Blind prediction exercise on modeling of PHWR fuel at extended burnup

    International Nuclear Information System (INIS)

    A blind prediction exercise was organised on Indian Pressurised Heavy Water Reactor (PHWR) fuel to investigate the predictive capability of existing codes for their application at extended burnup and to identify areas of improvement. The blind problem for this exercise was based on a PHWR fuel bundle irradiated in Kakrapar Atomic Power Station-I (KAPS-I) up to about 15 000 MWd/tU and subjected to detailed post-irradiation examination (PIE) in the hot cells facility at BARC. Eleven computer codes from seven countries participated in this exercise. The participants provided blind predictions of fuel temperature, fission gas release, internal gas pressure and other performance parameters for the fuel pins. The predictions were compared with the experimental PIE data which included fuel temperature derived from fuel restructuring, fission gas release measured by fuel pin puncturing, internal gas pressure in pin, cladding oxidation and fuel microstructural data. The details of the blind problem and an analysis of the results of blind predictions by the codes vis-a-vis measured data are provided in this paper

  13. Probabilistic assessment of dry transport with burnup credit

    International Nuclear Information System (INIS)

    The general concept of probabilistic analysis and its application to the use of burnup credit in spent fuel transport is explored. Discussion of the probabilistic analysis method is presented. The concepts of risk and its perception are introduced, and models are suggested for performing probability and risk estimates. The general probabilistic models are used for evaluating the application of burnup credit for dry spent nuclear fuel transport. Two basic cases are considered. The first addresses the question of the relative likelihood of exceeding an established criticality safety limit with and without burnup credit. The second examines the effect of using burnup credit on the overall risk for dry spent fuel transport. Using reasoned arguments and related failure probability and consequence data analysis is performed to estimate the risks of using burnup credit for dry transport of spent nuclear fuel. (author)

  14. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  15. Technical limitations of nuclear fuel materials and structures

    International Nuclear Information System (INIS)

    This report gives a summary of the tasks carried out within the project 'Technical limitations of nuclear fuel materials and structures' which belongs to the Finnish national research programme called 'Systems behaviour and operational aspects of safety'. The duration of the project was three years from 1990 to 1992. Most western LWR utilities, including the two Finnish ones have an incentive to implement extended burnup fuel cycles in their nuclear power plants. The aim of this project has been authorities to support them in the assessment and licensing of new fuel designs and materials. The research work of the project was focused on collecting and qualifying fuel performance data and on performing laboratory tests on fresh and irradiated cladding and structural materials. Moreover, knowledge of the high burnup phenomena was obtained through participation in international research projects such as OECD Halden Project and several Studsvik projects. Experimental work within the framework of the VVER fuel cooperative effort was also continued. (orig.)

  16. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data

  17. Post-irradiation examination of uranium-molybdenum dispersion fuel irradiated to high burn-up in NRU

    International Nuclear Information System (INIS)

    UMo dispersion fuels are promising candidates for research and test reactors. Mini-elements containing U7Mo and U10Mo (7 and 10 wt% Mo in U alloy) fuel particles dispersed in aluminium have been fabricated with a nominal loading of 4.5 gU/cm3. In order to compare the performance of the different UMo alloys, the mini-elements were irradiated adjacent to each other under nominally identical conditions in the National Research Universal (NRU) reactor. Maximum element linear ratings up to 100 kW/m and discharge burnups up to 80 atom% 235U were achieved. The experiment was conducted in phases such that adjacent pairs of mini-elements could be removed for post-irradiation examinations (PIE) after 20, 40, 60 and 80 atom% 235U burnup. PIE included underwater inspections, visual examinations and photography in the hot cells, gamma spectroscopy, dimensional measurements, immersion density measurements, metallography, and chemical burnup analysis. The results from the high burnup fuels are presented in this paper. The assessments compare the microstructural changes, porosity formation and fuel swelling in the two UMo dispersion fuels. The results indicate that U7Mo fuel is less stable that U10 Mo fuel under the conditions tested in NRU. (author)

  18. The US department of energy's transportation burnup credit program

    International Nuclear Information System (INIS)

    Aspects of the U. S. Department of Energy's (DOE's) transportation burnup credit program, the Department's motivation for conducting the program, and the status of burnup credit activities are presented. The benefits, technical, and regulatory considerations associated with using burnup credit for transport of irradiated nuclear fuel are discussed. The methods used in the DOE's actinide-only topical report are described in terms of the technical and regulatory issues. (authors)

  19. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    International Nuclear Information System (INIS)

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation

  20. Fission gas release behavior in high burnup UO{sub 2} fuels with developed rim-structure

    Energy Technology Data Exchange (ETDEWEB)

    Une, Katsumi [Global Nuclear Fuel-Japan Co. Led., Oarai, Ibaraki (Japan); Kashibe, Shinji [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Hayashi, Kimio [Japan Atomic Energy Research Inst., Tokyo (Japan)

    2002-11-01

    The effect of rim structure formation and external restraint pressure on fission gas release at transient conditions has been examined by using an out-of-pile high pressure heating technique for high burnup UO{sub 2} fuels (60, 74 and 90 GWd/t), which had been irradiated in test reactors. The latter two fuels bore a developed rim structure. The maximum heating temperature was 1500 degC, and the external pressures were independently controlled in the range of 10-150 MPa. The present high burnup fuel data were compared with those of previously studied BWR fuels of 37 and 54 GWd/t with almost no rim structure. The fission gas release and bubble swelling due to the growth of grain boundary bubbles and coarsened rim bubbles were effectively suppressed by the strong restraint pressure of 150 MPa for all the fuels; however the fission gas release remarkably increased for the two high burnup fuels with the developed rim structure, even at the strong restraint conditions. From the stepwise de-pressurization tests at an isothermal condition of 1500degC, the critical external pressure, below which a large burst release due to the rapid growth and interlinkage of the bubbles abruptly begins, was increased from a 40-60 MPa level for the middle burnup fuels to a high level of 120-140 MPa for the rim-structured high burnup fuels. The high potential for transient fission gas release and bubble swelling in the rim-structured fuels was attributed to highly over-pressurized fission gases in the rim bubbles. (author)

  1. Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

    International Nuclear Information System (INIS)

    The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies

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

    International Nuclear Information System (INIS)

    During irradiation, uranium dioxide (UO2) fuel undergo important restructuring mainly represented by densification and swelling, void migration, equiaxed grain growth, grain subdivision, and the formation of columnar grains. The purpose of this study is to obtain a comprehensive picture of the phenomenon of equiaxed grain growth in UO2 ceramic material. The change of the grain size in high-density uranium dioxide as a function of temperature, initial grain size, time, and burnup is calculated. Algorithm of fission gas release from UO2 fuel during high temperature irradiation at high burnup taking into account grain growth effect is presented. Theoretical results are compared with experimental data. (author)

  3. Modeling CANDU-type fuel behaviour during extended burnup irradiations using a revised version of the ELESIM code

    International Nuclear Information System (INIS)

    The high burnup database for CANDU fuel includes several cases from both power station and experimental reactor irradiations, with achieved burnups of up to 800 MW.h/kgU. The power history for each of these cases is different, encompassing low steady-state, declining, and power-ramps. This variety offers a good opportunity to check the models of fuel behaviour, and to identify areas for improvement. The main parameters for comparing calculated versus measured data are the fission gas release and the sheath hoop strain. Good agreement of calculated values of these two parameters with experimental data indicates that the global behaviour of the fuel element is adequately simulated by our codes. The ELESIM computer code was used as the simulation tool. The models for fission gas release, swelling and for fuel pellet expansion were thoroughly analysed. Changes were proposed for both models. The fuel pellet expansion model was modified to account for gaseous swelling, which becomes very important at high burnups. As well, the mathematics of the fission gas release model was upgraded for the diffusional release of fission gas atoms to the grain boundaries. A revised version of the ELESIM computer code was used to simulate the cases from the high burnup database. Satisfactory agreement was found for most cases. The discrepancies are discussed in view of alternative mechanisms that can operate and be enhanced at high burnup. These include stoichiometry changes with burnup that affects fission gas release, and also outer pellet rim fission gas release by a grain boundary diffusion process. The main conclusion of this study is that the revised version of the ELESIM code is able to simulate with reasonable accuracy high burnup as well as low burnup CANDU fuel. This includes irradiations of steady-state, declining, or ramped fuel power histories with a prolonged hold at high power. However, future improvements to ELESIM are needed to model fuel power histories with short dwell

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

    International Nuclear Information System (INIS)

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt% gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 μm in thickness after burnups of 11 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes, i.e., (1) partial amorphization of large Zr-Fe-Cr and Zr-Fe-Ni precipitates (300 to 800 nm in size), (2) virtually complete amorphization of small intermetallic precipitates and subsequent dissolution of the alloying elements, and (3) spinodal-like fluctuation and redistribution of the alloying elements following the amorphization and dissolution. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conducive to nodular oxidation. Besides the microstructural features associated with the intermetallic precipitates, TEM stereo electron microscopy revealed microscopic zirconium hydrides (30 to 100 nm in size) that were too small to be resolved by OM or SEM. Stereoscopic examination revealed a tendency for precipitation of the microscopic hydrides with c-component dislocations as the burnup increased. Also, an examination of bright- and dark-field stereopair images revealed three-dimensional distributions of fine cubic-zirconium-oxide precipitates (5 to 10 nm in size) and unidentified ''black-dot'' (5 to 10 nm) and ''white

  5. Local fission gas release from high burnup water reactor fuel under transient conditions

    International Nuclear Information System (INIS)

    The paper presents results for local gas release, produced by power transients (bump tests) at the end of life of a water reactor fuel. The burnup was from 2.7-4.5% FIMA (25,000-41,000 MWd/tU). The local linear power at the bump terminal level (BTL) of the fuel examined ranged from 300-415 W/cm. The hold time at BTL was either 24 or 72 h. Around 410-415 W/cm, the local gas releases measured on pellet sized samples were 35-40%. Radial xenon release profiles measured by electron microprobe analysis showed that onset of release occurred at about 700 deg. C. Above 1100 deg. C, a constant release of about 95% was found. (author)

  6. The role of ORIGEN-S in the design of burnup credit spent fuel casks

    International Nuclear Information System (INIS)

    Current licensing practices for spent fuel pools, storage facilities, and transportation casks require a conservative fresh fuel assumption be used in the criticality analysis. The U.S. Department of Energy is currently sponsoring a program to develop analysis methodologies and establish a new generation of spent fuel casks using the principle of burnup credit. The key difference in this new approach is the necessity to accurately predict the isotopic composition of the spent fuel. ORIGEN-S was selected to satisfy this requirement because of the flexibility and user-friendly input offered via its usage in the Standardized Computer Analyses for Licensing and Evaluation (SCALE) code system. This paper describes the fundamental role fulfilled by ORIGEN-S in the development of the analysis methodology, validation of the methods, definition of criticality safety margins and other licensing considerations in the design of a new generation of spent fuel casks. Particular emphasis is given to the performance of ORIGEN-S in comparisons with measurements of irradiated fuel compositions and in predicting isotopics for use in the calculation of reactor restart critical configurations that are performed as a part of the validation process

  7. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To obtain a nuclear fuel assembly having a function of eliminating corrosion products exfoliating from the surface of a fuel can, thereby reduce the radioactive crud in primary sodium coolant during operation of a FBR type reactor. Constitution: Nickel plates or grids made of metal plate with a nickel coated on the surface thereof are inserted in the upper blanket of a nuclear fuel element and between nuclear fuel element corresponding to the gas plenum. The nickel becomes helpful at high temperature in adsorbing Mn-54 which accounts for a major portion of the corrosion products. (J.P.N.)

  8. Main programs and techniques for examination of behaviour of the WWER high-burnup fuel in the MIR reactor

    International Nuclear Information System (INIS)

    This paper presents the main programs and techniques for examination of behaviour of the WWER high-burnup fuel performed in the MIR reactor as well as features of these techniques and some experimental results. These techniques were shown to provide examination under operating conditions including normal, transient and accidental conditions. Performance of the WWER standard fuel under the above various operating conditions is simulated by conducting different tests including repeated irradiation and transient tests of full-size and refabricated fuel rods as well as tests of the refabricated fuel rods under design-basis accidental (LOCA and RIA type) conditions and also tests of defective fuel rods. A high neutron flux density and heat removal conditions allow the performance of experiments with fuel having a burnup of ∼ 50 - 80 MWd/kgU at a linear power of ∼50 - 100 kW/m. The following types of rigs are used for testing: special irradiation rigs are used for testing under repeated power cycling and impulse conditions, instrumented fragments of WWER fuel assemblies containing refabricated high burnup fuel rods are also prepared for testing under design-basis accidental conditions (LOCA type). The experimental data presented in this paper can be used to check the conformity of the WWER fuel with licensing requirements in terms of the majority of criteria, as well as to estimate radiation consequences in case of cladding failure, to check and improve of computer codes for assessment of fuel rod states

  9. Nuclear fuel assembly spacer

    International Nuclear Information System (INIS)

    In a fuel assembly for a nuclear reactor a fuel element spacer formed of an array of laterally positioned cojoined tubular ferrules each providing a passage for one of the fuel elements, the elements being laterally supported in the ferrules between slender spring members and laterally oriented rigid stops

  10. A preliminary neutronic evaluation of the high temperature nuclear reactor (HTTR) using reprocessed fuel

    International Nuclear Information System (INIS)

    Highlights: • The HTTR was simulated using reprocessed fuels spiked with thorium and depleted uranium. • The effective neutron multiplication factor and the nuclear fuel evolution during the burn-up were analyzed. • The results indicated that reprocessed fuels can be used in the HTTR. - Abstract: The High Temperature Engineering Test Reactor (HTTR), a 30 MWth, graphite-moderated, helium-cooled reactor constructed by the Japanese government was simulated using reprocessed fuel obtained by UREX+ and spiked with thorium-232 and with depleted uranium. The effective neutron multiplication factor and the nuclear fuel evolution during the burn-up were analyzed. This study was performed using the ORNL SCALE 6.0 code, with CSAS6 and TRITON6 control modules. The results show in a preliminary way that the burn-up of reprocessed fuels in the HTTR core is possible, although the fissile material quantities should be increased while compared with the enrichments of the standard fuel

  11. Issues for effective implementation of burnup credit

    International Nuclear Information System (INIS)

    In the United States, burnup credit has been used in the criticality safety evaluation for storage pools at pressurized water reactors (PWRs) and considerable work has been performed to lay the foundation for use of burnup credit in dry storage and transport cask applications and permanent disposal applications. Many of the technical issues related to the basic physics phenomena and parameters of importance are similar in each of these applications. However, the nuclear fuel cycle in the United States has never been fully integrated and the implementation of burnup credit to each of these applications is dependent somewhat on the specific safety bases developed over the history of each operational area. This paper will briefly review the implementation status of burnup credit for each application area and explore some of the remaining issues associated with effective implementation of burnup credit. (author)

  12. Transient fission gas release from UO2 fuel for high temperature and high burnup

    International Nuclear Information System (INIS)

    In the present paper it is assumed that the fission gas release kinetics from an irradiated UO2 fuel for high temperature is determined by the kinetics of grain growth. A well founded assumption that Vitanza curve describes the change of uranium dioxide re-crystallization temperature and the experimental results referring to the limiting grain size presented in the literature are used to modify the grain growth model. Algorithms of fission gas release due to re-crystallization of uranium dioxide grains are worked out. The defect trap model of fission gas behaviour described in the earlier papers is supplemented with the algorithms. Calculations of fission gas release in function of time, temperature, burn-up and initial grain sizes are obtained. Computation of transient fission gas release in the paper is limited to the case where steady state of irradiation to accumulate a desired burn-up is performed below the temperature of re-crystallization then the subsequent step temperature increase follows. There are considered two kinds of step temperature increase for different burn-up: the final temperature of the step increase is below and above the re-crystallization temperature. Calculations show that bursts of fission gas are predicted in both kinds. The release rate of gas liberated for the final temperature above the re-crystallization temperature is much higher than for final temperature below the re-crystallization temperature. The time required for the burst to subside is longer due to grain growth than due to diffusion of bubbles and knock-out release. The theoretical results explain qualitatively the experimental data but some of them need to be verified since this sort of experimental data are not found in the available literature. (author)

  13. World nuclear fuel cycle requirements, 1984

    International Nuclear Information System (INIS)

    This report presents projections of the domestic and foreign requirements for uranium and enrichment services, as well as spent nuclear fuel discharges. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity published in a recent Energy Information Administration (EIA) report. Four scenarios (high, middle, low, and no new reactor orders) are included for domestic nuclear power capacity and three (high, middle, and low) for countries in the World Outside Planned Economies (WOCA). In addition, 4 sensitivity cases are presented for the US lower capacity factors, reactor aging, lower tails assay, and higher burnup. Six sensitivity cases are analyzed for the WOCA countries: (1) stable, instead of improving, capacity factors for the United States and for countries in the Other country group; (2) reactor aging; (3) recycling of uranium but not plutonium from spent fuel (the three standard scenarios assume recycling of both uranium and plutonium; (4) no recycling of spent fuels; (5) lower uranium enrichment tails assay; and (6) higher fuel burnup levels. The annual US requirements for uranium and for uranium enrichment service are projected to more than double between 1985 and 2020 in the middle case, and the cumulative amount of spent fuel discharged is projected to increase approximately 10-fold. Annual uranium requirements for the WOCA nations are projected to increase by about 60% between 1985 and 2000. In contrast, a 7- to 8-fold increase in U3O8 and enrichment service requirements is projected for the Other WOCA country group during this time period, as its relatively small existing nuclear power capacity undergoes rapid expansion

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  15. Preparation of uranium-plutonium carbide-based fuels simulating high burnup by carbothermic reduction and their properties

    International Nuclear Information System (INIS)

    Three types, hypostoichiometric, nearly stoichiometric and hyperstoichiometric, of uranium-plutonium carbide fuels simulating 10 at.% burnup were prepared by carbothermic reduction of oxide containing fission product elements. The carbides contained fission product phases such as the UMoC2 and the U2RuC2 type or the RECsub(1.5-2.0) phases (RE:rare earth). Composite theoretical densities of heterogenious carbides containing the UC, U2C3 type and fission product phases were calculated from the proportions and densities of these phases. By comparison of specific volume of the carbide between of 0 at.% and 10 at.% burnup, the solid fission product swelling rate of a carbide-based fuel was estimated to be 0.4-0.5 % per at.% burnup. (author)

  16. Influence of high burnup on the decay heat power of spent fuel at long-term storage

    International Nuclear Information System (INIS)

    Development and application of advanced fuel with higher burnup is now in practice of NPP with light water reactors in an increasing number of countries. High burnup allows to decrease significantly consumption of uranium. However, spent fuel of this type contains increased amount of high active actinides and fission products in comparison with spent fuel of common-type burnup. Therefore extended time of storage, improved cooling system of the storage facility will be required along with more strong radiation protection during storage, transportation and processing. Calculated data on decay heat power of spent uranium fuel of light water VVER-1000 type reactor are discussed in the paper. Long-term storage of discharged fuel during 100000 years is considered. Calculations were made for burnups of 40-70 MW d/kg. In the initial 50-year period of storage, power of fission products is much higher than that of actinides. Power of gamma-radiation is mainly due to fission products. During subsequent storage power of fission products quickly decreases, the main contribution to the power is given by actinides rather than by fission products. (author)

  17. Determination of plutonium’s isotopic composition in PWR spent nuclear fuel

    International Nuclear Information System (INIS)

    In this article the isotopic composition of the plutonium contained in spent nuclear fuel at discharge is analysed. The composition is analysed as a function of burn-up of a light water reactor using once-through cycle. The results are obtained using two different software products. Key words: pressurized water reactors, plutonium, spent nuclear fuel

  18. Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

    International Nuclear Information System (INIS)

    Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm3 by the fuel of density 4.00 gU/cm3 then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm3 in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29-62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1

  19. Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad 45650 (Pakistan); Aslam [Department of Physics and Applied Mathematics, PIEAS, Islamabad 45650 (Pakistan)]. E-mail: aslam_mcmaster@yahoo.com; Ahmad, Nasir [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad 45650 (Pakistan)

    2005-07-15

    Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm{sup 3} by the fuel of density 4.00 gU/cm{sup 3} then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm{sup 3} in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29-62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1.

  20. Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC)

    International Nuclear Information System (INIS)

    This report presents the results of the Coordinated Research Project (CRP) on Optimization of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plants (FUWAC, 2006-2009). It provides an overview of the results of the investigations into the current state of water chemistry practice and concerns in the primary circuit of water cooled power reactors including: corrosion of primary circuit materials; deposit composition and thickness on the fuel; crud induced power shift; fuel oxide growth and thickness; radioactivity buildup in the reactor coolant system (RCS). The FUWAC CRP is a follow-up to the DAWAC CRP (Data Processing Technologies and Diagnostics for Water Chemistry and Corrosion Control in Nuclear Power Plants 2001-2005). The DAWAC project improved the data processing technologies and diagnostics for water chemistry and corrosion control in nuclear power plants (NPPs). With the improved methods for controlling and monitoring water chemistry now available, it was felt that a review of the principles of water chemistry management should be undertaken in the light of new materials, more onerous operating conditions, emergent issues such as CIPS, also known as axial offset anomaly (AOA) and the ageing of operating power plant. In the framework of this CRP, water chemistry specialists from 16 nuclear utilities and research organizations, representing 15 countries, exchanged experimental and operational data, models and insights into water chemistry management. The CD-ROM attached to this IAEA-TECDOC includes the report itself, detailed progress reports of three Research Coordination Meetings (RCMs) (Annexes I-III) and the reports and presentations made during the project by the participants.

  1. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  2. Impact of the fission yield nuclear data uncertainties in the pin-cell burn-up OECD/NEA UAM Benchmark

    International Nuclear Information System (INIS)

    The prediction of fission products and the impact of their uncertainties to different safety-related spent fuel applications (burn-up credit, decay heat generation, radiological safety, waste management, burn-up prediction) are required for the evaluation of spent fuel system designs and safety analysis options. One of the nuclear data needs to this prediction is the independent fission yields. The mostly used general-purpose evaluated nuclear data libraries provide these data including their uncertainties as standard deviation, with no-correlation between fission yields. However, new developments in the theory and measurements of fission product yields are expected to result in new evaluated files in the next coming years. These files will include considerably more accurate yields including neutron energy dependence combined with new covariance information allowing realistic uncertainty estimates. In this paper, we focused on the effect of fission yield covariance information on criticality and depletion calculations. A LWR pin-cell burn-up benchmark, proposed in the general framework of the OECD/UAM Benchmark is analyzed to address the impact of independent fission yield uncertainties. Calculations were performed with the SCALE6 system and the ENDF/B-VII.1 fission yield data library, adding covariance data obtained from including covariance info of mass yields. Results are compared with those obtained with the uncertainty data currently provided by ENDF/B-VII.1. The uncertainty quantification is performed with a Monte Carlo sampling and then compared with linear perturbation. (author)

  3. Depleting a CANDU-6 fuel assembly using detailed burnup data and reactionwise energy release

    International Nuclear Information System (INIS)

    Temporal behavior of reactor fuel assembly due to neutron exposure is an integral part of lattice analysis. It is important to estimate the production of actinides and fission products as a function of burnup so as to decide the quality of fuel for further energy production. It is also important from the point of view of post irradiation behavior of fuel. The information on heat production during and after irradiation helps in determining the amount of time a fuel assembly needs to be cooled before taking it up for storage or reprocessing. In the present study we have considered the CANDU-6 fuel assembly as reference. Lattice analysis has been performed using development version of code DRAGON. A total of 192 nuclides have been selected as part of the analysis, of which 19 are actinides, 151 are fission products and the rest are structural elements. The fission products have been treated explicitly. There is no pseudo fission product. Using DRAGR module, a multigroup microscopic cross section library in DRAGLIB format has been generated. An important aspect of this library is the explicit treatment of most neutron induced reactions. We have for the first time attempted to perform power normalization due to energy from various neutron induced reactions including (n, γ), (n, f), (n, 2n), (n, 3n), (n, 4n), (n, α), (n, p), (n, 2α), (n, np), (n, d), (n, t). Energy due to decay has also been considered explicitly. Even though the decay energy contributes very little relative to the neutron induced reactions, the information will be very useful for post irradiation behavior of fuel. It was observed that the maximum contributing reactions for the power normalization are (n, f), (n, γ) and (n, 2n). We have assessed the contribution of fission products and actinides towards power normalization as a function of burnup. We have also studied the pinwise contribution towards power normalization in each ring of CANDU-6 fuel. We have attempted to compare the effect of

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

    International Nuclear Information System (INIS)

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

  5. Evolution of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear fuel is the primary energy source for sustaining the nuclear fission chain reactions in a reactor. The fuels in the reactor cores are exposed to highly aggressive environment and varieties of advanced fuel materials with improved nuclear properties are continuously being developed to have optimum performance in the existing core conditions. Fabrications of varieties of nuclear fuels used in diverse forms of reactors are mainly based on two naturally occurring nuclear source elements, uranium as fissile 235U and fertile 238U, and thorium as fertile 232Th species. The two metals in the forms of alloys with specific elements, ceramic oxides like MOX and ceramic non-oxide as mixed carbide and nitride with suitable nuclear properties like higher metal density, thermal conductivity, etc. are used as fuels in different reactor designs. In addition, efficiency of various advanced fuels in the forms of dispersion, molten salt and other types are also under investigations. The countries which have large deposits of thorium but limited reserves of uranium, are trying to give special impetus on the development of thorium-based fuels for both thermal and fast reactors in harnessing nuclear energy for peaceful uses of atomic energy. (author)

  6. Burnup Credit of French PWR-MOx fuels: methodology and associated conservatisms with the JEFF-3.1.1 evaluation

    International Nuclear Information System (INIS)

    Considering spent fuel management (storage, transport and reprocessing), the approach using 'fresh fuel assumption' in criticality-safety studies results in a significant conservatism in the calculated value of the system reactivity. The concept of Burnup Credit (BUC) consists in considering the reduction of the spent fuel reactivity due to its burnup. A careful BUC methodology, developed by CEA in association with AREVA-NC was recently validated and written up for PWR-UOx fuels. However, 22 of 58 French reactors use MOx fuel, so more and more irradiated MOx fuels have to be stored and transported. As a result, why industrial partners are interested in this concept is because taking into account this BUC concept would enable for example a load increase in several fuel cycle devices. Recent publications and discussions within the French BUC Working Group highlight the current interest of the BUC concept in PWR-MOx spent fuel industrial applications. In this case of PWR-MOx fuel, studies show in particular that the 15 FPs selected thanks to their properties (absorbing, stable, non-gaseous) are responsible for more than a half of the total reactivity credit and 80% of the FPs credit. That is why, in order to get a conservative and physically realistic value of the application keff and meet the Upper Safety Limit constraint, calculation biases on these 15 FPs inventory and individual reactivity worth should be considered in a criticality-safety approach. In this context, thanks to an exhaustive literature study, PWR-MOx fuels particularities have been identified and by following a rigorous approach, a validated and physically representative BUC methodology, adapted to this type of fuel has been proposed, allowing to take fission products into account and to determine the biases related to considered isotopes inventory and to reactivity worth. This approach consists of the following studies: - isotopic correction factors determination to guarantee the criticality

  7. Burn-up effect on instant release from an initial corrosion of UO2 and MOX fuel under anoxic conditions

    International Nuclear Information System (INIS)

    The objective of the work is to obtain instant release experimental values for different radionuclides as a function of spent fuel type (UO2 and MOX) and burn-up (from 30 to 63 MWd/kgU) that will be useful for the performance assessment studies related to the behaviour of spent fuel under repository conditions or, in any case, spent fuel conditions in which labile radionuclides can be released. To determine the instant release source terms, sets of leaching experiments were conducted with spent UO2 and MOX fuel with burnups ranging from 30 to 63 MWd/kg U in presence of cladding as the container material. The fuels were leached in carbonated groundwater (CW) having a buffered pH of 7.5 at room temperature. Some observations are also made of the differences in matrix dissolution behaviour of the different fuels based on observed U, Pu and Np concentrations The ultimate issue is to evaluate the differences in the ''instant'' inventory measurements for spent fuels in order to provide experimental data that allow to evaluate the source terms used in the safety-assessment calculations, and to improve the accuracy of such data for the future. It is important to remark that the quality of the experimental results obtained describes the influence of the spent fuel (SF) burn-up on fast release of inventory fraction (release under 200 days). (authors)

  8. Advances in fuel pellet technology for improved performance at high burnup. Proceedings of a Technical Committee meeting

    International Nuclear Information System (INIS)

    The IAEA has recently completed two co-ordinated Research Programmes (CRPs) on The Development of Computer Models for Fuel Element Behaviour in Water Reactors, and on Fuel Modelling at Extended Burnup. Through these CRPs it became evident that there was a need to obtain data on fuel behaviour at high burnup. Data related o thermal behaviour, fission gas release and pellet to clad mechanical interaction were obtained and presented at the Technical Committee Meeting on Advances in Fuel Pellet Technology for Improved Performance at High Burnup which was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). The 34 papers from 10 countries are published in this proceedings and presented by a separate abstract. The papers were grouped in 6 sessions. First two sessions covered the fabrication of both UO2 fuel and additives and MOX fuel. Sessions 3 and 4 covered the thermal behaviour of both types of fuel. The remaining two sessions dealt with fission gas release and the mechanical aspects of pellet to clad interaction

  9. Quality and Reliability Aspects in Nuclear Power Reactor Fuel Engineering

    International Nuclear Information System (INIS)

    In order to decrease costs and increase competitiveness, nuclear utilities use more challenging operational conditions, longer fuel cycles and higher burnups, which require modifications in fuel designs and materials. Different aspects of quality assurance and control, as well as analysis of fuel performance have been considered in a number of specialized publications. The present publication provides a concise but comprehensive overview of all interconnected quality and reliability issues in fuel fabrication, design and operation. It jointly tackles technical, safety and organizational aspects, and contains examples of state of the art developments and good practices of coordinated work of fuel designers, vendors and reactor operators

  10. Checking of the spent fuel assemblies burnup based on the results of the neutron flux measuring using IAEA fork detector at Zaporozhye NPP, Units 1, 2, 4 and 6

    International Nuclear Information System (INIS)

    Burnup calculation for fuel assemblies are made for simulation of a fuel cycle checking the following experimental data: power curve; absorber rods position during the cycle; temperature of cooling water on the active area input; cooling water consumption and boric acid concentration. The burnup is calculated for ten areas along the fuel assembly. The method for the burnup checking is presented

  11. Stereological evolution of the rim structure in PWR-fuels at prolonged irradiation: Dependencies with burn-up and temperature

    International Nuclear Information System (INIS)

    The stereology of the rim-structure was studied for PWR-fuels up to the ninth irradiation cycle, achieving maximum local burn-ups of 240 GWd/tM and beyond. At intermediate radial positions (0.55 0 c = 0.29. Rim-cavities are expected to remain closed at least up to this limit

  12. Study of morphology on the oxidation and the annealing of high burn-up UO2 spent fuel

    International Nuclear Information System (INIS)

    The morphology of the high burnup UO2 spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent UO2 fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at 1400 .deg. C for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to 500 .deg. C and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the UO2 matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to 10 μm

  13. Study of morphology on the oxidation and the annealing of high burn-up UO{sub 2} spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Jae Geun; Yang, Yong Sik; Song, Keun Woo; Lee, Hyung Kwon; Kwon, Hyung Moon [KAERI, Daejon (Korea, Republic of)

    2005-12-15

    The morphology of the high burnup UO{sub 2} spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent UO{sub 2} fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at 1400 .deg. C for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to 500 .deg. C and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the UO{sub 2} matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to 10 {mu}m.

  14. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    CERN Document Server

    Chiesa, Davide; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2015-01-01

    A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate experimental reactors from power ones, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the...

  15. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to publish