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Sample records for axial burnup profile

  1. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    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)

  2. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

    International Nuclear Information System (INIS)

    Wagner, J.C.; DeHart, M.D.

    2000-01-01

    This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ''end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified

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

    International Nuclear Information System (INIS)

    Bznuni, S.

    2006-01-01

    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 k eff 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 k eff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect

  4. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  5. ZZ PWR-AXBUPRO-GKN, Measured Axial Burnup Profiles, NPP Neckarewstheim

    International Nuclear Information System (INIS)

    Neuber, Jens-Christian; Lamprecht, Thomas

    1999-01-01

    Description or function: PWR-AXBUPRO-GKN12 contains Axial Burnup Shapes released by Siemens AG Power Generation Group. It contains data sets relative to following NPP and initial enrichment: - NPP Neckarwestheim 1, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.8 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 4.0 wt.-% 235-U. In each of these files the axial shapes are listed one after the other. Each shape is characterised by: - the number of the cycle; - the number of the fuel assembly. - axial burnup shape characteristics: - height (in centimeter) of the nodes with respect to the active length of the fuel assemblies normalized to the cold, unirradiated state. - the nodal burnup (in MWd/kg U). - fuel assembly design data as well as core geometry and operating data pertinent to depletion calculations: NPP Neckarwestheim 1 (GKN1) - (square pitch lattice 15X15 - thermal Power 2497 MW)thermal Power 2497 MW); More than 700 EOC axial shapes from cycle 18 up. From cycle 18 to cycle 20 a change from an Out-In-Loading to an In-Out-Loading has taken place. Fuel assemblies up to number 1093 have spacer grids made of Inconel, whereas all the fuel assemblies from number 1094 up have spacer grids made of Zircaloy. Discharge burnups range from: 9.7 to 52.8 MWd/kg. NPP Neckarwestheim 2 (GKN2) - (square pitch lattice 18X18 - thermal Power 3850 MW) more than 500 EOC axial shapes from cycle 5 up: - More than 170 shapes for an initial fuel enrichment of 3.5 wt.-% 235-U, discharge burnup ranges from 16.3 to 44.4 MWd/kg; - more than 170 shapes for an initial fuel enrichment of 3.8 wt.-% 235-U, discharge burnup ranges from 14.0 to 52.8 MWd/kg; - more than 180 shapes for an initial fuel enrichment of 4.0 wt.-% 235-U. discharge burnup ranges from 15.5 to 48.9 MWd/kg. PWR AXBUPRO

  6. Conservative axial burnup distributions for actinide-only burnup credit

    International Nuclear Information System (INIS)

    Kang, C.; Lancaster, D.

    1997-11-01

    Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model each burnup mesh with the corresponding isotopic compositions. Although this approach represents a significant increase in modeling efforts compared to the uniform average burnup approach, it can adequately determine the reactivity effect of the axial burnup distribution. A major consideration is what axial burnup distributions are appropriate for use in light of many possible distributions depending on core operating conditions and histories. This paper summarizes criticality analyses performed to determine conservative axial burnup distributions. The conservative axial burnup distributions presented in this paper are included in the Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages, Revision 1 submitted in May 1997 by the US Department of Energy (DOE) to the US Nuclear Regulatory Commission (NRC). When approved by NRC, the conservative axial burnup distributions may be used to model PWR spent nuclear fuel for the purpose of gaining actinide only burnup credit

  7. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  8. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    International Nuclear Information System (INIS)

    Barkauskas, V.; Plukiene, R.; Plukis, A.

    2016-01-01

    Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

  9. Determination of axial profit performed burnup credit by SCALE 4.3-system

    International Nuclear Information System (INIS)

    Miro, R.; Verdu, G.; Munoz-Cobo, J. L.

    1998-01-01

    SCALE 4.3 is a modular code system designed for realizing standard computational analysis for licensing evaluation. Since now, spent fuel storage pools criticality analysis have been done considering this fuel as fresh, with its maximum enrichment. With burnup credit we can obtain cheaper and compact configurations. The procedure for calculating a spent fuel storage consists of a burnup calculation plus a criticality calculation. We can perform a conservative approximation for the burnup calculations using 1-D results, but, besides the geometry configurations for the 3-D criticality calculation. we need an appropriate approximation to model the burnup axial variation. We assume that for a burnup profile set, the most conservative profile is between the lower and the upper range of this profile, set. We consider only combinations of the maximum and minimum burnup in each axial region, for each burnup range. This gives an estimation of the different burnup shapes effect and the general characteristics of the most conservative shape. (Author) 6 refs

  10. The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

    International Nuclear Information System (INIS)

    Kilger, R.; Gmal, B.; Moser, E.F.

    2008-01-01

    Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

  11. Triton burnup in JET - profile effects

    International Nuclear Information System (INIS)

    Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van

    1991-01-01

    Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small ( 2 /s). (author) 4 refs., 3 figs

  12. Triton burnup in JET - profile effects

    Energy Technology Data Exchange (ETDEWEB)

    Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van (Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking); Adams, J.M.; Watkins, N. (AEA Industrial Technology, Harwell Laboratory (United Kingdom))

    1991-01-01

    Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small (<<0.1 m[sup 2]/s). (author) 4 refs., 3 figs.

  13. Burn-Up Determination by High Resolution Gamma Spectrometry: Axial and Diametral Scanning Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Blackadder, W H; Ronqvist, N

    1967-02-15

    In the gamma spectrometric determination of burn-up the use of a single fission product as a monitor of the specimen fission rate is subject to errors caused by activity saturation or, in certain cases, fission product migration. Results are presented of experiments in which all the resolvable gamma peaks in the fission product spectrum have been used to calculate the fission rate; these results form a pattern which reflect errors in the literature values of the gamma branching ratios, fission yields etc., and also represent a series of empirical correction factors. Axial and diametral scanning experiments on a long-irradiated low-enrichment fuel element are also described and demonstrate that it is possible to differentiate between fissions in U-235 and in Pu-239 respectively by means of the ratios of the Ru-106 activity to the activities of the other fission products.

  14. Original paper Cytokine profiles in axial spondyloarthritis

    Directory of Open Access Journals (Sweden)

    Marta Madej

    2015-04-01

    Full Text Available Objectives: Current studies concentrate on the cytokine network and its role in the pathogenesis of spondyloarthritis (SpA. In this study, we analyzed whether the serum cytokine profile (interleukins: IL-10, IL-11, IL-12, IL-15, IL-17, IL-23 and IL-33 correlates with demographic data, clinical manifestations, disease activity and treatment outcome in a group of patients with axial spondyloarthritis. Material and methods: Forty-nine patients with an established diagnosis of axial spondyloarthritis (aSpA and 19 healthy volunteers as controls were enrolled in the study. Clinical evaluation included patient’s medical history, 44 joint count, back pain intensity and global disease activity in the preceding week (VAS, the duration of morning stiffness and blood tests. Disease activity was assessed using BASDAI and ASDAS-CRP. Serum concentration of IL-10, IL-11, IL-12, IL-15, IL-17, IL-23 and IL-33 was determined. Results : In patients with aSpA, elevated serum concentration of IL-10, IL-15, IL-17 and IL-23 was detected. In the aSpA group we detected higher values of serum concentration of IL-23 and IL-33 in the subgroup with anterior uveitis (83.1 ±184.0 pg/ml vs. 14.0 ±17.1 pg/ml, p < 0.0001 and 45.5 ±71.9 pg/ml vs. 18.4 ±14.3 pg/ml, p < 0.0001, respectively. Additionally, in the subgroup with peripheral arthritis, elevation of serum concentration of IL-12 (249.3 ±246.9 pg/ml vs. 99.9 ±105.9 pg/ml, p = 0.0001 was detected. Patients with preradiological SpA had higher serum concentration of IL-17 than patients with established diagnosis of AS (6.37 ±8.50 pg/ml vs. 2.04 ±2.98 pg/ml, p = 0.0295. No differences in serum concentration of analyzed cytokines were found between the subgroup with low to moderate disease activity and the subgroup with high to very high disease activity. Conclusions : We report that in aSpA patients, compared to controls, elevated serum concentrations of IL-10, IL-15, IL-17 and IL-23 were observed. Some cytokines may

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

    International Nuclear Information System (INIS)

    BSC

    2004-01-01

    , initial 235 U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed

  16. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

    2008-01-01

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  17. Axial Compressive Strength of Foamcrete with Different Profiles and Dimensions

    Directory of Open Access Journals (Sweden)

    Othuman Mydin M.A.

    2014-01-01

    Full Text Available Lightweight foamcrete is a versatile material; primarily consist of a cement based mortar mixed with at least 20% volume of air. High flow ability, lower self-weight, minimal requirement of aggregate, controlled low strength and good thermal insulation properties are a few characteristics of foamcrete. Its dry densities, typically, is below 1600kg/m3 with compressive strengths maximum of 15MPa. The ASTM standard provision specifies a correction factor for concrete strengths of between 14 and 42MPa to compensate for the reduced strength when the aspect height-to-diameter ratio of specimen is less than 2.0, while the CEB-FIP provision specifically mentions the ratio of 150 x 300mm cylinder strength to 150 mm cube strength. However, both provisions requirements do not specifically clarify the applicability and/or modification of the correction factors for the compressive strength of foamcrete. This proposed laboratory work is intended to study the effect of different dimensions and profiles on the axial compressive strength of concrete. Specimens of various dimensions and profiles are cast with square and circular cross-sections i.e., cubes, prisms and cylinders, and to investigate their behavior in compression strength at 7 and 28 days. Hypothetically, compressive strength will decrease with the increase of concrete specimen dimension and concrete specimen with cube profile would yield comparable compressive strength to cylinder (100 x 100 x 100mm cube to 100dia x 200mm cylinder.

  18. Evaluation of efficiency of axial profiling in WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Ananjev, Yu. A.; Kurakin, K. Yu.; Artemov, V.G.; Ivanov, A.S.

    2005-01-01

    The present report deals with consideration of fuel enrichment axial profiling in WWER-440 assemblies. The study is performed on improving the effectiveness of fuel utilization using the example of implementing the axial profiling in the assemblies of the second generation. For simulation of fuel loadings the computer code package SAPFIR 9 5 and RC is used that allows for correct consideration of specific features of assemblies design changes. The methodical approach to assessment of effectiveness of implementing the axial profiling is considered with the use of capabilities of the mentioned code package. In conclusion the recommendations are given on using the fuel enrichment axial profiling in WWER-440 assemblies (Authors)

  19. COGEMA/TRANSNUCLEAIRE's experience with burnup credit

    International Nuclear Information System (INIS)

    Chanzy, Y.; Guillou, E.

    1998-01-01

    Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)

  20. Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

    2001-11-01

    Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

  1. EVOLUT - a computer program for fast burnup evaluation

    International Nuclear Information System (INIS)

    Craciunescu, T.; Dobrin, R.; Stamatescu, L.; Alexa, A.

    1999-01-01

    EVOLUT is a computer program for burnup evaluation. The input data consist on the one hand of axial and radial gamma-scanning profiles (for the experimental evaluation of the number of nuclei of a fission product - the burnup monitor - at the end of irradiation) and on the other hand of the history of irradiation (the time length and values proportional to the neutron flux for each step of irradiation). Using the equation of evolution of the burnup monitor the flux values are iteratively adjusted, by a multiplier factor, until the calculated number of nuclei is equal to the experimental one. The flux values are used in the equation of evolution of the fissile and fertile nuclei to determine the fission number and consequently the burnup. EVOLUT was successfully used in the analysis of several hundreds of CANDU and TRIGA-type fuel rods. We appreciate that EVOLUT is a useful tool in the burnup evaluation based on gamma spectrometry measurements. EVOLUT can be used on an usual AT computer and in this case the results are obtained in a few minutes. It has an original and user-friendly graphical interface and it provides also output in script MATLAB files for graphical representation and further numerical analysis. The computer program needs simple data and it is valuable especially when a large number of burnup analyses are required quickly. (authors)

  2. Current studies related to the use of burnup credit in France

    International Nuclear Information System (INIS)

    Raby, Jerome; Lavarenne, Caroline; Barreau, Anne; Riffard, Cecile; Roque, Benedicte; Bioux, Philippe; Doucet, Michel; Guillou, Eric; Leka, Georges; Toubon, Herve

    2003-01-01

    In order to avoid criticality risks, a large number of facilities using spent fuels have been designed considering the fuel as fresh. This choice has obviously led to considerable safety margins. In the early 80's, a method was accepted by the French Safety Authorities allowing to consider the changes in the fuel composition during the depletion with some very pessimistic hypothesis: only actinides were considered and the amount of burnup used in the studies was equal to the mean burnup in the 50-least-irradiated centimeters. As many facilities still want to optimize their processes (e.g. transportation, storage, fuel reprocessing), the main companies involved in the French nuclear industry, researchers and IRSN set up a Working Group in order to study the way burnup could be taken into account in the criticality calculations, considering some fission products and a more realistic axial profile of burnup. The first of this article introduces the current French method used to take burnup into account in the criticality studies. The second part is devoted to the studies achieved by the Working Group to improve this method, especially concerning the consideration of the neutron absorption of some fission products and of an axial profile of burnup: for that purpose, some results are presented related to the steps of the process like the depletion calculations, the definition of an axial profile and the criticality calculation. In the third part, some results (keff) obtained with fission products and an axial profile are compared to those obtained with the current one. The conclusions presented are related to the present state of knowledge and may differ from the final conclusions of the Working Group. (author)

  3. Investigation of Vacuum Arc Voltage Characteristics Under Different Axial Magnetic Field Profiles

    International Nuclear Information System (INIS)

    Jia Shenli; Song Xiaochuan; Huo Xintao; Shi Zongqian; Wang Lijun

    2010-01-01

    Characteristics of the arc voltage under different profiles of axial magnetic field were investigated experimentally in a detachable vacuum chamber with five pairs of specially designed electrodes generating both bell-shaped and saddle-shaped magnetic field profile. The arc column and cathode spot images were photographed by a high speed digital camera. The dependence of the arc voltage on arcing evolution is analyzed. It is indicated that the axial magnetic field profile could affect the arc behaviors significantly, and the arc voltage is closely related to the arc light intensity.

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

    International Nuclear Information System (INIS)

    Neuber, J.C.

    1998-01-01

    The paper describes the experience gained in Germany in applying burnup credit methodologies to wet storage and dry transport systems of spent LWR fuel. It gives a survey of the levels of burnup credit presently used or intended to be used, the regulatory status and future developments planned, the codes used for performing depletion and criticality calculations, the methods applied to verification of these codes, and the methods used to treat parameters specific of burnup credit. In particular it is shown that the effect of axial burnup profiles on wet PWR storage designs based on burnup credit varies from fuel type to fuel type. For wet BWR storage systems the method of estimating a loading curve is described which provides for a given BWR fuel assembly design the minimum required initial burnable absorber content as a function of the initial enrichment of the fuel. (author)

  5. Technical development on burn-up credit for spent LWR fuels

    International Nuclear Information System (INIS)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

    2000-10-01

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

  6. 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.

  7. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

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

  8. The Sensitivity Analysis of Axial Pressure Tube Creep Profile for Dryout Power in PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Euiseung; Kim, Youngae [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Stern Laboratory performed the CHF tests with only one axial pressure tube creep profile per 3.3%, 5.1% peak crept channel and made CHF correlation including creep factor from the CHF test results. Wolsong nuclear power plants also have utilized the same CHF correlation derived by CNL. Pressure tube diameter creep rate is function of fast neutron, coolant temperature, and coolant pressure in a channel. It means that various axial pressure tube creep profiles exist in PHWR due to the history of operating conditions. Usually, CHF correlation is used during ROP(Regional Overpower Protection) Trip Setpoint Analysis or Safety Analysis in PHWR. The sensitivity analysis for CHF effects using various creep profiles is needed. This paper summarizes the comparison results of dryout power between CHF test creep profile and estimated creep profiles of Wolsong units. The effect of axial pressure tube creep profile for dryout power in fuel channel is evaluated by using Stern Lab. CHF test creep profile and 380 channel creep profiles of Wolsong. The dryout powers at 3.3% and 5.1% test conditions are slightly smaller when using 380 Wolsong channels creep profiles. These also show that the simulated dryout powers maintain consistency regardless of flow conditions.

  9. An analytical calculation of the axial density profile for 1-d slab expansion

    International Nuclear Information System (INIS)

    Ho, D

    1999-01-01

    Obtaining an analytical expression for the axial density profile can provide us with a quick and convenient way to evaluate the density evolution for targets with different densities and dimensions. In this note, we show that such an analytical expression can be obtained based on the self-similar solutions and the method of characteristics for 1-D slab expansion

  10. Dose profile measurement in computerized axial tomography equipment using thermoluminescent dosemeters

    International Nuclear Information System (INIS)

    Azorin V, J.C.; Falcony, C.; Azorin N, J.

    2000-01-01

    In this work are presented the results about measuring the radiation dose profile in two equipment of computerized axial tomography (Tac). Thermoluminescent dosemeters (Dtl) of LiF, Mg, Cu, P + Ptfe in form of disks were used which were developed and made in Mexico. The results showed that Dtl are appropriated for these type of studies. (Author)

  11. 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)

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-01-01

    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

  12. Detection of uranium extraction zone by axial temperature profiles in a pulsed column for Purex process

    International Nuclear Information System (INIS)

    Tsukada, T.; Takahashi, K.

    1991-01-01

    A new method was presented for detecting uranium extraction zone in a pulsed column by means of measuring axial temperature profile originated from reaction heat during uranium extraction. Key parameters of the temperature profiles were estimated with a code developed for calculating temperature profiles in a direct-contact heat exchanger such as a pulsed column, and were verified using data from a small pulsed column simulating reaction heat with injecting hot water. Finally, the results were compared with those from an actual uranium extraction tests, indicating that the method presented was promising for detecting uranium extraction zone in a pulsed column. (author)

  13. Isocrit: a burnup credit tool for spent fuel pool storage calculations - 333

    International Nuclear Information System (INIS)

    Kucukboyaci, V.N.; Marshall, W.J.

    2010-01-01

    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 up-rate, exit temperature changes, etc) with a quick turnaround. (authors)

  14. Experimental investigation of an annular diffuser for axial fans at different inflow profiles

    Directory of Open Access Journals (Sweden)

    Walter Johannes

    2017-01-01

    Full Text Available Axial fans are used in power plants for fresh air supply and flue gas transport. A typical configuration consists of an axial fan and annular diffuser which connects the fan to the following piping. In order to achieve a high efficiency of the con-figuration, not only the components have to be optimized but also their interaction. The present study focuses on the diffuser of the configuration. Experiments are performed on a diffuser-piping configuration to investigate the influence of the velocity profile at the fan outlet on the pressure recovery of the configuration. Two different diffuser inlet profiles are generated, an undisturbed profile and a profile with the typical outlet characteristics of a fan. The latter is generated by the superposition of screens in the inlet zone. The tests are conducted at a high Reynolds number (Re ≈ 4∙105. Mean velocity profiles and wall shear stresses are measured with hydraulic methods (Prandtl and Preston tubes. The results show that there is a lack of momentum at the outer wall of the diffuser and high shear stresses at the inner wall in case of the undisturbed inflow profile. For the typical fan outlet profile it is vice versa. There are high wall shear stresses at the outer wall while the boundary layer of the inner wall lacks momentum. The pressure recovery of the undisturbed inflow configuration is in good agreement with other studies.

  15. PLUTON: A Three-Group Model for the Radial Distribution of Plutonium, Burnup, and Power Profiles in Highly Irradiated LWR Fuel Rods

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe

    2001-01-01

    A three-group model (PLUTON) is described, which predicts the power density distribution, plutonium buildup, and burnup profiles across the fuel pellet radius as a function of in-pile time and parameters characterizing the type of reactor system with respect to fuel temperature and changes of density during the irradiation period. The PLUTON model is a part of two fuel performance codes (ASFAD and FEMAXI-V), which provide all necessary input for this model, mainly local temperatures and fuel matrix density across the radius. Comparisons between measurements and predictions of the PLUTON model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnup between 21 000 and 64 000 MWd/t. It is shown that the PLUTON predictions are in good agreement with measurements as well as with predictions of the well-known TUBRNP model. The proposed model is flexibly applicable for all types of light water reactor (LWR) fuels, including mixed oxide, and for fuel tested in the Organization for Economic Corporation and Development's Halden heavy water reactor. The PLUTON three-group model is based on analytical (theoretical) consideration of neutron absorption in a resonant region of the fuel in its apparent form. It makes the model more flexible in comparison with the semi-empirical TUBRNP one-group model and allows the physically based model analysis of commercial LWR-type fuels at high burnup as well as analysis of experimental fuel rods tested in the Halden heavy water reactor, which is one of the main test reactors in the world. The differences in fuel behavior in the Halden reactor in terms of burnup distribution and plutonium buildup can be more clearly understood with the PLUTON model

  16. Influence of an axial magnetic field on the density profile of capillary plasma channels

    CERN Document Server

    Ivanov, V V; Toma, E S; Bijkerk, F

    2003-01-01

    A narrow capillary plasma channel, with a sizeable depletion of the electron density on the channel axis, has been proposed to guide a laser pulse over a length of several to several tens of centimetres. We discuss the possibility to significantly improve the wave-guiding properties of such a channel by applying an axial magnetic field. Our analytical and numerical studies show that a pulsed axial magnetic field of 10 T in a hydrogen capillary plasma at a pressure of 50 Torr will reduce the on-axis plasma density by a factor of three, and the full width at half maximum of the density profile by a factor of two. The resulting parabolic plasma density profile is expected to be more efficient in guiding laser pulses.

  17. Mathematical description of tooth flank surface of globoidal worm gear with straight axial tooth profile

    Science.gov (United States)

    Połowniak, Piotr; Sobolak, Mariusz

    2017-12-01

    In this article, a mathematical description of tooth flank surface of the globoidal worm and worm wheel generated by the hourglass worm hob with straight tooth axial profile is presented. The kinematic system of globoidal worm gear is shown. The equation of globoid helix and tooth axial profile of worm is derived to determine worm tooth surface. Based on the equation of meshing the contact lines are obtained. The mathematical description of globoidal worm wheel tooth flank is performed on the basis of contact lines and generating the tooth side by the extreme cutting edge of worm hob. The presented mathematical model of tooth flank of TA worm and worm wheel can be used e.g. to analyse the contact pattern of the gear.

  18. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

  19. Burnup credit for storage and transportation casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1988-01-01

    The application of burnup credit to storage and transportation cask licensing results in a significant improvement in cask capacity and an associated reduction of the cost per kilogram of uranium in the cask contents. The issues for licensing with burnup credit deal primarily with the treatment of fission product poisons and methods of verification of burnup during cask operations. Other issues include benchmarking of cross-section sets and codes and the effect of spatial variation of burnup within an assembly. The licensing of burnup credit for casks will be complex, although the criticality calculations are not themselves difficult. Attention should be directed to the use of fission product poisons and the uncertainties that they introduce. Verification of burnup by measurements will remove some of the concerns for criticality safety. Calculations for burnup credit casks should consider rod-to-rod and axial variations of burnup, as well as variability of burnable poisons it they are used in the assembly. In spite of the complexity of cask burnup credit licensing issues, these issues appear to be resolvable within the current state of the art of criticality safety

  20. Application of routine methods for the inspector fuel burn-up determination and identification of displacement of spent fuel elements by dummy elements

    International Nuclear Information System (INIS)

    Rohar, S.

    1979-08-01

    14 irradiated assemblies were analyzed using nondestructive high resolution gamma spectrometry (HRGS). Measured and calculated (on the basis of calorimetric data) axial burnup profiles and average burnup values were compared. The measurements of spent fuel were performed in the Bohunice A-1 dry hot cell by using a proper collimating system and the standard Agency equipment, consisting of PGT intrinsic Ge detectors and Silena MCA with 1024 channels. The method of 134 Cs/ 137 Cs fission product activity ratio was used for burnup determination. It was found that the burnup values for 14 measured assemblies determined by HRGS were systematically lower than the calculated values with about 4-5%. The difference between the nondestructively determined burnup value of the 2N0053 assembly (average over 11 measured points) and destructively determined burnup (average over 19 measured points) was less than 2%. Passive neutron measurements of the irradiated assembly showed that the neutron counting rate was high enough for practical use and that the neutron and gamma profiles were similar and close to the burnup profile. Some calculations of gamma ray activity angular distribution were made for different numbers of dummy elements inside the irradiated assemblies. The results show that, by using gamma spectrometry transversal method, it is possible to find a significant number of dummy elements in different types of assemblies

  1. High burnup issues and modelling strategies

    International Nuclear Information System (INIS)

    Dutta, B.K.

    2005-01-01

    The performance of high burnup fuel is affected by a number of phenomena, such as, conductivity degradation, modified radial flux profile, fission gas release from high burnup structures, PCMI, burnup dependent thermo-mechanical properties, etc. The modelling strategies of some of these phenomena are available in literature. These can be readily incorporated in a fuel modelling performance code. The computer code FAIR has been developed in BARC over the years to evaluate the fuel performance at extended burnup and modelling of the fuel rods for advanced fuel cycles. The present paper deals with the high burnup issues in the fuel pins, their modelling strategies and results of the case studies specifically involving high burnup fuel. (author)

  2. Measuring Density Profiles of Electrons and Heavy Particles in a Stable Axially Blown Arc

    Science.gov (United States)

    Carstensen, J.; Stoller, P.; Galletti, B.; Doiron, C. B.; Sokolov, A.

    2017-08-01

    Two-color spatial carrier wave interferometry employing pulsed 532- and 671-nm lasers is used to measure the electron-density and heavy-particle-density profiles in the stagnation point of a stable, axially blown arc in argon for currents of 50 to 200 A and stagnation point pressures of 0.2 to 16 bar. This technique takes advantage of the fact that the free-electron contribution to the refractive index depends strongly on the wavelength, while that of the heavy particles does not. The high spatial resolution achieved allows the hot core of the arc to be readily distinguished from the surrounding boundary layer. A custom-built test device is used to ensure flow conditions that lead to a stable, axisymmetric arc; this permits the reconstruction of the density and temperature profiles using a single projection (interferometric image) of the refractive-index distribution through the arc (at two wavelengths). The arc radius determined from the heavy-particle density decreases with increasing stagnation pressure and increases with the current. These measurements are in good agreement with a simple axially blown arc model taking into account Ohmic heating, radiation losses, and enthalpy flow for core temperatures of approximately 16 500 K. The measured electron density at the center of the arc agrees well with a prediction based on local thermodynamic equilibrium.

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

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    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 (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. 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 k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  4. Influence of axial movement on fatigue of ProFile Ni-Ti rotary instruments: an in vitro evaluation.

    Science.gov (United States)

    Avoaka, Marie-Chantal; Haïkel, Youssef

    2010-05-01

    The aim of this study was to evaluate the influence of the axial movement and the angle of curve (in degrees) on fatigue of nickel-titanium (Ni-Ti) ProFile rotary endodontic instruments. Ni-Ti ProFile rotary instruments (Maillefer SA, Ballaigues, Switzerland), 25 mm long in the range of ISO size 15 to 40 with two tapers (0.4 and 0.6) were evaluated. They are divided in two groups: the instruments with axial movement and those without axial movement. The system used to test the fatigue is maintained in mechanical conditions as close as possible to the clinical situation. The axial movement is in the order of 2 mm in corono-apical direction with a frequency of 1 Hz. The concave radii incorporating a notched V-form for guiding the instruments were: 5; 7,5 and 10 mm. The rotary system is mounted on an electric handpiece and rotated at 350 rpm speed as recommended by the manufacturers. The instruments are rotated until their separation, and the time, in seconds, is recorded. Statistical evaluation is undertaken using a two-way t-test to identify significant differences between variables in the study (p engine drive ProFile instruments incorporating an axial movement and the instruments without axial movement with the same radius of curvature, size and taper.The incorporation of the axial movement increases significantly the life-span of the ProFile rotary instruments. This should reduce the risk of the instrument separation during the endodontic treatment.

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

    International Nuclear Information System (INIS)

    DOE

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

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

  7. Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

    International Nuclear Information System (INIS)

    Košťál, Michal; Švadlenková, Marie; Baroň, Petr; Rypar, Vojtěch; Milčák, Ján

    2016-01-01

    Highlights: • Fission density in partly flooded compact core. • Calculation of fission density axial profile. • Significant calculational under prediction of experimental axial profile. - Abstract: Measurement and calculation of the axial power profile near the boundary of a moderated and non-moderated core is used to analyze the suitability of the neutron-physical process description, mainly the angular cross-section of a water-moderated uranium system. This is also an important issue because it affects the radiation situation above the partly flooded core of a water-moderated reactor. Axial power profiles of various fuel pins irradiated on reactor LR-0 were measured and the results were compared with MCNP6 code calculations using the ENDF/B-VII.0 nuclear data library. The calculated power profile in positions above the moderator level significantly underestimates experimental results. This might be caused by an improper description of the angular distribution of scattered neutrons in a water-moderated uranium system.

  8. Choosing the optimum burnup

    International Nuclear Information System (INIS)

    Geller, L.; Goldstein, L.; Franks, W.A.

    1986-01-01

    This paper reviews some of the considerations utilities must evaluate when going to higher discharge burnups. The advantages and disadvantages of higher discharge burnups are described, as well as a consistent approach for evaluating optimum discharge burnup and its comparison to current practice. When an analysis is performed over the life of the plant, the design of the terminal cycles has significant impact on the lifetime savings from higher burnups. Designs for high burnup cycles have a greater average inventory value in the core. As one goes to higher burnup, there is a greater likelihood of discarding a larger value in unused fuel unless the terminal cycles are designed carefully. This effect can be large enough in some cases to wipe out the lifetime cost savings relative to operating with a higher discharge burnup cycle

  9. 'CANDLE' burnup regime after LWR regime

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nagata, Akito

    2008-01-01

    CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

  10. Velocity profile of water vapor inside a cavity with two axial inlets and two outlets

    Science.gov (United States)

    Guadarrama-Cetina, José; Ruiz Chavarría, Gerardo

    2014-03-01

    To study the dynamics of Breath Figure phenomenon, a control of both the rate of flow and temperature of water vapor is required. The experimental setup widely used is a non hermetically closed chamber with cylindrical geometry and axial inlets and outlets. In this work we present measurements in a cylindrical chamber with diameter 10 cm and 1.5 cm height, keeping a constant temperature (10 °C). We are focused in the velocity field when a gradient of the temperatures is produced between the base plate and the vapor. With a flux of water vapor of 250 mil/min at room temperature (21 °C), the Reynolds number measured in one inlet is 755. Otherwise, the temperatures of water vapor varies from 21 to 40 °C. The velocity profile is obtained by hot wire anemometry. We identify the stagnations and the possibly instabilities regions for an empty plate and with a well defined shape obstacle as a fashion sample. Facultad de Ciencias, UNAM.

  11. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

    Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics

  12. Profiling Cognitive Deficits in Intra-Axial and Extra-Axial Tumors Using Addenbrooke's Cognitive Examination as a Screening Tool: An Indian Experience.

    Science.gov (United States)

    Cherkil, Sandhya; Panikar, Dilip; Soman, Deepak Kuttikkattu

    2017-01-01

    Tumors of the brain, whether intra- or extra-axial, results in cognitive deficits. The aim of the present study was to profile cognitive deficits using Addenbrooke's Cognitive Examination-Malayalam (ACE-M) as a screen and to determine the sensitivity and specificity of the same. Seventy-four drug naïve patients diagnosed to have brain tumors were assessed for cognitive functioning using ACE-M before surgery. Patients with high-grade intra-axial tumors showed a significant association on the cognitive domains of registration (0.04), recall (0.01), and visuospatial functioning (0.02). Gender showed an association between registration (0.02) and verbal fluency (0.02) with females performing better while education was significantly associated with retrograde or remote memory (0.00) with college-educated sample performing better. Significance was assumed at P cognitive decline on the cognitive domains of attention (0.02), recall (0.05), naming (0.02), and language functions (0.01). College educated group performed better on registration (0.01), recall (0.09), naming (0.00), and visuospatial functioning (0.00). The area under the receiver operating characteristic curve was estimated as 0.75, which indicates fairly good discriminative ability with a cut off of 71/100; sensitivity at 77.3 and specificity fixed at 67. ACE-M is capable of bringing out cognitive deficits along with a number of cognitive domains in patients with intra- and extra-axial tumors in the capacity of a screen, with fairly good levels of sensitivity and specificity.

  13. Physical models for high burnup fuel

    International Nuclear Information System (INIS)

    Kanyukova, V.; Khoruzhii, O.; Likhanskii, V.; Solodovnikov, G.; Sorokin, A.

    2003-01-01

    In this paper some models of processes in high burnup fuel developed in Src of Russia Troitsk Institute for Innovation and Fusion Research are presented. The emphasis is on the description of the degradation of the fuel heat conductivity, radial profiles of the burnup and the plutonium accumulation, restructuring of the pellet rim, mechanical pellet-cladding interaction. The results demonstrate the possibility of rather accurate description of the behaviour of the fuel of high burnup on the base of simplified models in frame of the fuel performance code if the models are physically ground. The development of such models requires the performance of the detailed physical analysis to serve as a test for a correct choice of allowable simplifications. This approach was applied in the SRC of Russia TRINITI to develop a set of models for the WWER fuel resulting in high reliability of predictions in simulation of the high burnup fuel

  14. End effect Keff bias curve for actinide-only burnup credit casks

    International Nuclear Information System (INIS)

    Kang, C.H.; Lancaster, D.B.

    1997-01-01

    A conservative end effect k eff bias curve for actinide-only burnup credit for spent fuel casks is presented in this paper. The k eff bias values can be added to the uniform axial burnup analysis to conservatively bound the actinide-only end effect. A normalized axial burnup distribution for the standard Westinghouse 17 x 17 assembly design is used for calculating k eff . The end effect calculated is a strong function of burnup, and increases as cask size size decreases. The presence of poison plates increases the end effect. The bias curve presented is based on the most limiting cask configuration of a single PWR assembly with completely black poison plates. Therefore, axially uniform criticality calculations with application of the proposed k eff could eliminate the need for axially burnup dependent analyses. 7 refs., 1 fig

  15. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    International Nuclear Information System (INIS)

    2011-01-01

    compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.

  16. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Enercon Services, Inc.

    2011-03-14

    compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.

  17. High Burnup Effects Program

    International Nuclear Information System (INIS)

    Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D.

    1990-04-01

    This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs

  18. Appropriate burnup measurements for transportation burnup credit

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.

    1997-01-01

    This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy requirements are established for dependent and independent systems. The requirements have been tested and are achievable, ensuring safe operation without extra cost. 6 refs

  19. Application of Candle burnup to small fast reactor

    International Nuclear Information System (INIS)

    Sekimoto, H.; Satoshi, T.

    2004-01-01

    A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. An equilibrium state was obtained for a large fast reactor (core radius is 2 m and reflector thickness is 0.5 m) successfully by using a newly developed direct analysis code. However, it is difficult to apply this burnup strategy to small reactors, since its neutron leakage becomes large and neutron economy becomes worse. Fuel enrichment should be increased in order to sustain the criticality. However, higher enrichment of fresh fuel makes the CANDLE burnup difficult. We try to find some small reactor designs, which can realize the CANDLE burnup. We have successfully find a design, which is not the CANDLE burnup in the strict meaning, but satisfies qualitatively its characteristics mentioned at the top of this abstract. In the final paper, the general description of CANDLE burnup and some results on the obtained small fast reactor design are presented.(author)

  20. Burnup credit in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    2001-01-01

    The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analytical activities in the country are described, as well as international collaborations in this field. Perspectives for future research and development of new applications are finally resumed. (author)

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

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    2005-11-01

    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)

  2. End effects in the criticality analysis of burnup credit casks

    International Nuclear Information System (INIS)

    Brady, M.C.; Parks, C.V.

    1990-01-01

    A study to evaluate the effect of axially dependent burnup on k eff has been performed as part of an effort to qualify procedures to be used in establishing burnup credit in shipping cask design and certification. This study was performed using a generic 31-element modular cast-iron cask (wall thickness 33.1 cm) with a 1-cm-thick borated stainless-steel basket for reactivity control. Fuel isotopics used here are those of the 17 x 17 Westinghouse assemblies from the North Anna Unit 1 reactor. Virginia Power (VP) provided detailed spatial isotopics for the fuel assemblies in-core at beginning-of-cycle 5 (BOC-5) as generated from their PDQ analyses. Twenty-two axial planes were defined in the original VP data. The isotopics used in this study were for a 3.41 initial wt % 235 U and an average burnup of 31.5 GWd/MTU

  3. The influence of the tangential velocity of inner rotating wall on axial velocity profile of flow through vertical annular pipe with rotating inner surface

    Directory of Open Access Journals (Sweden)

    Sharf Abdusalam M.

    2014-03-01

    Full Text Available In the oil and gas industries, understanding the behaviour of a flow through an annulus gap in a vertical position, whose outer wall is stationary whilst the inner wall rotates, is a significantly important issue in drilling wells. The main emphasis is placed on experimental (using an available rig and computational (employing CFD software investigations into the effects of the rotation speed of the inner pipe on the axial velocity profiles. The measured axial velocity profiles, in the cases of low axial flow, show that the axial velocity is influenced by the rotation speed of the inner pipe in the region of almost 33% of the annulus near the inner pipe, and influenced inversely in the rest of the annulus. The position of the maximum axial velocity is shifted from the centre to be nearer the inner pipe, by increasing the rotation speed. However, in the case of higher flow, as the rotation speed increases, the axial velocity is reduced and the position of the maximum axial velocity is skewed towards the centre of the annulus. There is a reduction of the swirl velocity corresponding to the rise of the volumetric flow rate.

  4. The Effects of Wear upon the Axial Profile of a Grinding Wheel in the Construction of Innovative Grinding Wheels for Internal Cylindrical Grinding

    OpenAIRE

    Nadolny, K.; Słowiński, B.

    2011-01-01

    The article describes the effects of wear upon the axial profile of a grinding wheel in the axial cylindrical grinding processes. This mechanism was used to develop a grinding wheel with zone diversified structure made of microcrystalline sintered corundum abrasive grains and vitrifies bond. Such a grinding wheel is characterized by the conical rough grinding zone that is made by grains of a relatively large size, and a cylindrical finish grinding zone with grains of a smaller size and can be...

  5. Triton burnup in JET

    International Nuclear Information System (INIS)

    Chipsham, E.; Jarvis, O.N.; Sadler, G.

    1989-01-01

    Triton burnup measurements have been made at JET using time-integrated copper activation and time-resolved silicon detector techniques. The results confirm the classical nature of both the confinement and the slowing down of the 1 MeV tritons in a plasma. (author) 8 refs., 3 figs

  6. Evaluation of diagnostic efficiency of intelligence program for automatic diagnosis of circumferential profiles of myocardial short axial SPECT

    International Nuclear Information System (INIS)

    Bunko, Hisashi; Tada, Akira; Nakajima, Kenichi; Taki, Junichi; Kojima, Kazuhiko.

    1984-01-01

    Quantitative diagnosis using circumferential profile (CFP) is widely employed for Tl-201 myocardial imaging. However, its diagnostic efficiency is susceptible to the number and composition of the confirmed normal (CN) group. Intelligence program for CFP data file and automatic diagnosis (CAD), which was previously developed by us, was modified for creating universal data file (d-base) and evaluated its diagnostic efficiency for myocardial short axial images according to the number of CNs. CAD consists of following 3 major functions: (1) CFP data filing to d-base, (2) modification and correction of file data, and (3) automatic diagnosis of CFP. When function (1) or (2) is completed, new diagnostic criteria (mean-2 sd) are generated automatically using CN data in the d-base at that time. Because of this function, diagnostic criteria are changeable according to the number of CNs. Sensitivity (TP) and specificity (TN) of CAD program for Tl-201 7-pinhole images of 16 normals and 17 ischemic heart diseases (IHDs) were 100%, 44% (CN = 4), 88%, 94% (CN = 8 or 16) and 88%, 100% (CN = 12), respectively. Diagnostic efficiency reached plateau in more than 8 CNs. Although number of patients were limited (8 CNs and 6 IHDs), results of CAD for short axial SPECT showed consistent tendency with 7-pinhole images. In conclusion, intelligence program is necessary for automatic diagnosis using quantitative criteria under the clinical condition of daily increasing informations. Automatic diagnosis of CFP using CAD program is simple, effective and useful for interpretation of Tl-201 myocardial short axial images. CAD program is applicable to any quantitative distribution study using CFPs. (author)

  7. The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

    2003-03-01

    In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.

  8. The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

    International Nuclear Information System (INIS)

    Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

    2003-03-01

    In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc

  9. In-shoe plantar tri-axial stress profiles during maximum-effort cutting maneuvers.

    Science.gov (United States)

    Cong, Yan; Lam, Wing Kai; Cheung, Jason Tak-Man; Zhang, Ming

    2014-12-18

    Soft tissue injuries, such as anterior cruciate ligament rupture, ankle sprain and foot skin problems, frequently occur during cutting maneuvers. These injuries are often regarded as associated with abnormal joint torque and interfacial friction caused by excessive external and in-shoe shear forces. This study simultaneously investigated the dynamic in-shoe localized plantar pressure and shear stress during lateral shuffling and 45° sidestep cutting maneuvers. Tri-axial force transducers were affixed at the first and second metatarsal heads, lateral forefoot, and heel regions in the midsole of a basketball shoe. Seventeen basketball players executed both cutting maneuvers with maximum efforts. Lateral shuffling cutting had a larger mediolateral braking force than 45° sidestep cutting. This large braking force was concentrated at the first metatarsal head, as indicated by its maximum medial shear stress (312.2 ± 157.0 kPa). During propulsion phase, peak shear stress occurred at the second metatarsal head (271.3 ± 124.3 kPa). Compared with lateral shuffling cutting, 45° sidestep cutting produced larger peak propulsion shear stress (463.0 ± 272.6 kPa) but smaller peak braking shear stress (184.8 ± 181.7 kPa), of which both were found at the first metatarsal head. During both cutting maneuvers, maximum medial and posterior shear stress occurred at the first metatarsal head, whereas maximum pressure occurred at the second metatarsal head. The first and second metatarsal heads sustained relatively high pressure and shear stress and were expected to be susceptible to plantar tissue discomfort or injury. Due to different stress distribution, distinct pressure and shear cushioning mechanisms in basketball footwear might be considered over different foot regions. Copyright © 2014 Elsevier Ltd. All rights reserved.

  10. A compact CMA spectrometer with axially integrated hybrid electron-ion gun for ISS, AES and sputter depth profile analysis

    International Nuclear Information System (INIS)

    Gisler, E.; Bas, E.B.

    1986-01-01

    Until now, the combined application of electrons and ions in surface analysis required two separate sources for electrons and ions with different incidence angles. The newly developed hybrid electron-ion gun, however, allows bombardment of the same sample area both with noble gas ions and with electrons coming from the same direction. By integrating such a hybrid gun axially in a cylindrical mirror energy analyser (CMA) a sensitive compact single flange spectrometer obtains for ion scattering spectroscopy (ISS), Auger electron spectroscopy (AES), and sputtering all within normal beam incidence. This concept makes accurate beam centering very easy. Additionally, the bombardment from the same direction both for sputtering and for surface analysis brings advantages in depth profiling. The scattering angle for ISS has a constant value of about 138 0 . The hybrid gun delivers typically an electron beam current of -20μA at 3keV for AES, and an ion beam current of +40 nA and +1.2μA at 2 keV for ISS and sputtering respectively. The switching time between ISS, AES, and sputtering mode is about 0.1 s. So this system is best suited for automatically controlled depth profile analysis. The design and operation of this new system will be described and some applications will be discussed. (author)

  11. Numerical performance analysis of acoustic Doppler velocity profilers in the wake of an axial-flow marine hydrokinetic turbine

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, Marshall C.; Harding, Samuel F.; Romero Gomez, Pedro DJ

    2015-09-01

    The use of acoustic Doppler current profilers (ADCPs) for the characterization of flow conditions in the vicinity of both experimental and full scale marine hydrokinetic (MHK) turbines is becoming increasingly prevalent. The computation of a three dimensional velocity measurement from divergent acoustic beams requires the assumption that the flow conditions are homogeneous between all beams at a particular axial distance from the instrument. In the near wake of MHK devices, the mean fluid motion is observed to be highly spatially dependent as a result of torque generation and energy extraction. This paper examines the performance of ADCP measurements in such scenarios through the modelling of a virtual ADCP (VADCP) instrument in the velocity field in the wake of an MHK turbine resolved using unsteady computational fluid dynamics (CFD). This is achieved by sampling the CFD velocity field at equivalent locations to the sample bins of an ADCP and performing the coordinate transformation from beam coordinates to instrument coordinates and finally to global coordinates. The error in the mean velocity calculated by the VADCP relative to the reference velocity along the instrument axis is calculated for a range of instrument locations and orientations. The stream-wise velocity deficit and tangential swirl velocity caused by the rotor rotation lead to significant misrepresentation of the true flow velocity profiles by the VADCP, with the most significant errors in the transverse (cross-flow) velocity direction.

  12. The measurement of abundance and content of 148Nd monitor for the determination of burnup with mass spectrometry

    International Nuclear Information System (INIS)

    Gao Shuqin; Li Silin

    1992-09-01

    The abundance and content of nuclide 148 Nd, which is used as monitor to determine reactor element burnup, were measured by mass spectrometry, and the burnup can be calculated from measured results. The distribution of 148 Nd abundance and content in the axial direction are consistent with the theoretical calculation. The burnup values agree with the data obtained from heavy isotope ratio and radiochemistry methods within the errors of 4.0% and 2.8% respectively

  13. 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.

  14. ABB high burnup fuel

    International Nuclear Information System (INIS)

    Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

  15. Thomson scattering on argon surfatron plasmas at intermediate pressures: Axial profiles of the electron temperature and electron density

    International Nuclear Information System (INIS)

    Palomares, J.M.; Iordanova, E.; Veldhuizen, E.M. van; Baede, L.; Gamero, A.; Sola, A.; Mullen, J.J.A.M. van der

    2010-01-01

    The axial profiles of the electron density n e and electron temperature T e of argon surfatron plasmas in the pressure range of 6-20 mbar and microwave power between 32 and 82 W have been determined using Thomson Scattering of laser irradiation at 532 nm. For the electron density and temperature we found values in the ranges 5 x 10 18 e 19 m -3 and 1.1 e e and T e down to 8% and 3%, respectively. It is found that n e decreases in the direction of the wave propagation with a slope that is nearly constant. The slope depends on the pressure but not on the power. Just as predicted by theories we see that increasing the power leads to longer plasma columns. However, the plasmas are shorter than what is predicted by theories based on the assumption that for the plasma-wave interaction electron-atom collisions are of minor importance (the so-called collisionless regime). The plasma vanishes long before the critical value of the electron density is reached. In contrast to what is predicted by the positive column model it is found that T e does not stay constant along the column, but monotonically increases with the distance from the microwave launcher. Increases of more than 50% over 30 cm were found.

  16. Thomson scattering on argon surfatron plasmas at intermediate pressures: Axial profiles of the electron temperature and electron density

    Energy Technology Data Exchange (ETDEWEB)

    Palomares, J.M., E-mail: f02palij@gmail.co [Departamento de Fisica, Universidad de Cordoba, Campus de Rabanales, ed. C-2, 14071 Cordoba (Spain); Iordanova, E.; Veldhuizen, E.M. van; Baede, L. [Department of Applied Physics, Eindhoven University of Technology, PO Box 513, 5600 MB Eindhoven (Netherlands); Gamero, A.; Sola, A. [Departamento de Fisica, Universidad de Cordoba, Campus de Rabanales, ed. C-2, 14071 Cordoba (Spain); Mullen, J.J.A.M. van der, E-mail: j.j.a.m.v.d.Mullen@tue.n [Department of Applied Physics, Eindhoven University of Technology, PO Box 513, 5600 MB Eindhoven (Netherlands); Departamento de Fisica, Universidad de Cordoba, Campus de Rabanales, ed. C-2, 14071 Cordoba (Spain)

    2010-03-15

    The axial profiles of the electron density n{sub e} and electron temperature T{sub e} of argon surfatron plasmas in the pressure range of 6-20 mbar and microwave power between 32 and 82 W have been determined using Thomson Scattering of laser irradiation at 532 nm. For the electron density and temperature we found values in the ranges 5 x 10{sup 18} < n{sub e} < 8 x 10{sup 19} m{sup -3} and 1.1 < T{sub e} < 2.0 eV. Due to several improvements of the setup we could reduce the errors of n{sub e} and T{sub e} down to 8% and 3%, respectively. It is found that n{sub e} decreases in the direction of the wave propagation with a slope that is nearly constant. The slope depends on the pressure but not on the power. Just as predicted by theories we see that increasing the power leads to longer plasma columns. However, the plasmas are shorter than what is predicted by theories based on the assumption that for the plasma-wave interaction electron-atom collisions are of minor importance (the so-called collisionless regime). The plasma vanishes long before the critical value of the electron density is reached. In contrast to what is predicted by the positive column model it is found that T{sub e} does not stay constant along the column, but monotonically increases with the distance from the microwave launcher. Increases of more than 50% over 30 cm were found.

  17. Modelling the high burnup UO2 structure in LWR fuel

    International Nuclear Information System (INIS)

    Lassmann, K.; Walker, C.T.; Laar, J. van de; Lindstroem, F.

    1995-01-01

    The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U. (orig.)

  18. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    Ahlf, J.

    1983-01-01

    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.) [de

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

    International Nuclear Information System (INIS)

    1998-04-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

    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. Refs, figs, tabs.

  1. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

    2013-01-01

    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

  2. Research on burnup physics

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1974-07-01

    One of the major problems in burnup studies is the reasonably fast and accurate calculation of the space-and-energy dependent neutron flux and reaction rates for realistic power reactor fuel geometries and compositions, and its optimal integration in the global reactor calculations. The scope of the present research was to develop improved methods trying to satisfy the above requirements. In the epithermal region, simple and efficient approximation is proposed which allows the analytical solution for the space dependence of the spherical harmonics flux moments, and hence the derivation of the recurrence relations between he flux moments at successive lethargy pivotal points. A new matrix formalism to invert the coefficient matrix of band structure resulted in a reduce computer time and memory demands. The research on epithermal region is finalized in computing programme SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, and the related integral quantities as reaction rates and resonance integrals. For partial verification of the above methods a Monte Carlo procedure was developed. Using point-wise representation of variables, a flexible and fast convergent integral transport method SEPT i developed. Expanding the neutron source and flux in finite series of arbitrary polynomials, the space-and-energy dependent integral transport equation is transformed into a general linear algebraic form, which is solved numerically. A simple and efficient procedure for deriving multipoint equations and constructing matrix is proposed and examined, and no unwanted oscillations were noticed. The energy point method was combined with the spherical harmonics method as well. A multi zone few-group program SPECTAR for global reactor calculations was developed. For testing, the flux distribution, neutron leakage and effective multiplication factor for the PWR reactor of the power station San Onofre were calculated. In order to verify

  3. Influence of Axial Movement on Fatigue of Profile® Ni-Ti Rotary Instruments: an in vitro evaluation

    Directory of Open Access Journals (Sweden)

    Marie-Chantal Avoaka

    2010-05-01

    Full Text Available The aim of this study was to evaluate the influence of the axial movement and the angle of curve (in degrees on fatigue of nickel-titanium (Ni-Ti ProFile’ rotary endodontic instruments.Ni-Ti ProFile’ rotary instruments (Maillefer SA, Ballaigues, Switzerland, 25 mm long in the range of ISO size 15 to 40 with two tapers (0.4 and 0.6 were evaluated. They are divided in two groups: the instruments with axial movement and those without axial movement. The system used to test the fatigue is maintained in mechanical conditions as close as possible to the clinical situation. The axial movement is in the order of 2mm in corono-apical direction with a frequency of 1Hz. The concave radii incorporating a notched V-form for guiding the instruments were: 5; 7,5 and 10 mm. The rotary system is mounted on an electric handpiece and rotated at 350 rpm speed as recommended by the manufacturers. The instruments are rotated until their separation, and the time, in seconds, is recorded. Statístícal evaluation is undertaken using a two-way t-test to identify significant differences between variables in the study (p <0,05.We found significant statistical difference (p<0,05 between Ni-Ti engine drive ProFile’ instruments incorporating an axial movement and the instruments without axial movement with the same radius of curvature, size and taper.The incorporation of the axial movement increases significantly the life-span of the ProFile’ rotary instruments. This should reduce the risk of the instrument separation during the endodontic treatment

  4. Influence of axial movement on fatigue of PROFILE* NI-TI rotary instruments: an in vitro evaluation

    Science.gov (United States)

    Avoaka, Marie-Chantal; Haïkel, Youssef

    2010-01-01

    The aim of this study was to evaluate the influence of the axial movement and the angle of curve (in degrees) on fatigue of nickel-titanium (Ni-Ti) ProFile’ rotary endodontic instruments. Ni-Ti ProFile’ rotary instruments (Maillefer SA, Ballaigues, Switzerland), 25 mm long in the range of ISO size 15 to 40 with two tapers (0.4 and 0.6) were evaluated. They are divided in two groups: the instruments with axial movement and those without axial movement. The system used to test the fatigue is maintained in mechanical conditions as close as possible to the clinical situation. The axial movement is in the order of 2mm in corono-apical direction with a frequency of 1Hz. The concave radii incorporating a notched V-form for guiding the instruments were: 5; 7,5 and 10 mm. The rotary system is mounted on an electric handpiece and rotated at 350 rpm speed as recommended by the manufacturers. The instruments are rotated until their separation, and the time, in seconds, is recorded. Statístícal evaluation is undertaken using a two-way t-test to identify significant differences between variables in the study (p engine drive ProFile’ instruments incorporating an axial movement and the instruments without axial movement with the same radius of curvature, size and taper. The incorporation of the axial movement increases significantly the life-span of the ProFile’ rotary instruments. This should reduce the risk of the instrument separation during the endodontic treatment. PMID:20507289

  5. Optimization of TRU burnup in modular helium reactor

    International Nuclear Information System (INIS)

    Yonghee, Kim; Venneri, F.

    2007-01-01

    An optimization study of a single-pass TRU (transuranic) deep-burn (DB) has been performed for a block-type MHR (Modular Helium Reactor) proposed by General Atomics. Assuming a future equilibrium scenario of advanced LWRs, a high-burnup TRU vector is considered: 50 GWD/MTU and 5-year cooling. For 3-D equilibrium cores, the performance analysis is done by using a continuous energy Monte Carlo depletion code MCCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial block shuffling strategy in terms of the fuel burnup and core power distributions. The impact of the kernel size of TRISO fuel is evaluated and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of TRISO fuels. A higher graphite density is evaluated in terms of the fuel burnup. In addition, it is shown that the core power distribution can be effectively controlled by zoning of the packing fraction of TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a small batch size for fuel reloading, at the expense of a marginal decrease of the TRU discharge burnup. Depending on the fuel management scheme, fuel specifications, and core parameters, the TRU burnup in an optimized DB-MHR core is over 60% in a single-pass irradiation campaign. (authors)

  6. Improvements for Monte Carlo burnup calculation

    Energy Technology Data Exchange (ETDEWEB)

    Shenglong, Q.; Dong, Y.; Danrong, S.; Wei, L., E-mail: qiangshenglong@tsinghua.org.cn, E-mail: d.yao@npic.ac.cn, E-mail: songdr@npic.ac.cn, E-mail: luwei@npic.ac.cn [Nuclear Power Inst. of China, Cheng Du, Si Chuan (China)

    2015-07-01

    Monte Carlo burnup calculation is development trend of reactor physics, there would be a lot of work to be done for engineering applications. Based on Monte Carlo burnup code MOI, non-fuel burnup calculation methods and critical search suggestions will be mentioned in this paper. For non-fuel burnup, mixed burnup mode will improve the accuracy of burnup calculation and efficiency. For critical search of control rod position, a new method called ABN based on ABA which used by MC21 will be proposed for the first time in this paper. (author)

  7. The radial distribution of plutonium in high burnup UO2 fuels

    International Nuclear Information System (INIS)

    Lassmann, K.; O'Carroll, C.; Laar, J. van de; Walker, C.T.

    1994-01-01

    A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21 000 and 64 000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions. (orig.)

  8. Electron temperature profiles in axial field 2.45 GHz ECR ion source with a ceramic chamber

    Science.gov (United States)

    Abe, K.; Tamura, R.; Kasuya, T.; Wada, M.

    2017-08-01

    An array of electrostatic probes was arranged on the plasma electrode of a 2.45 GHz microwave driven axial magnetic filter field type negative hydrogen (H-) ion source to clarify the spatial plasma distribution near the electrode. The measured spatial distribution of electron temperature indicated the lower temperature near the extraction hole of the plasma electrode corresponding to the effectiveness of the axial magnetic filter field geometry. When the ratio of electron saturation current to the ion saturation current was plotted as a function of position, the obtained distribution showed a higher ratio near the hydrogen gas inlet through which ground state hydrogen molecules are injected into the source. Though the efficiency in producing H- ions is smaller for a 2.45 GHz source than a source operated at 14 GHz, it gives more volume to measure spatial distributions of various plasma parameters to understand fundamental processes that are influential on H- production in this type of ion sources.

  9. Studies on validation possibilities for computational codes for criticality and burnup calculations of boiling water reactor fuel; Untersuchungen zu Validierungsmoeglichkeiten von Rechencodes fuer Kritikalitaets- und Abbrandrechnungen von Siedewasserreaktor-Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthais; Hannstein, Volker; Kilger, Robert; Sommer, Fabian; Stuke, Maik

    2017-06-15

    The Application of the method of Burn-up Credit on Boiling Water Reactor fuel is much more complex than in the case of Pressurized Water Reactors due to the increased heterogeneity and complexity of the fuel assemblies. Strongly varying enrichments, complex fuel assembly geometries, partial length fuel rods, and strong axial variations of the moderator density make the verification of conservative irradiation conditions difficult. In this Report, it was investigated whether it is possible to take into account the burn-up in criticality analyses for systems with irradiated Boiling Water Reactor fuel on the basis of freely available experimental data and by additionally applying stochastic methods. In order to achieve this goal, existing methods for stochastic analysis were adapted and further developed in order to being applicable to the specific conditions needed in Boiling Water Reactor analysis. The aim was to gain first insight whether a workable scheme for using burn-up credit in Boiling Water Reactor applications can be derived. Due to the fact that the different relevant quantities, like e.g. moderator density and the axial power profile, are strongly correlated, the GRS-tool SUnCISTT for Monte-Carlo uncertainty quantification was used in the analysis. This tool was coupled to a simplified, consistent model for the irradiation conditions. In contrast to conventional methods, this approach allows to simultaneously analyze all involved effects.

  10. Power ramp tests of high burnup BWR segment rods

    International Nuclear Information System (INIS)

    Hayashi, H.; Etoh, Y.; Tsukuda, Y.; Shimada, S.; Sakurai, H.

    2002-01-01

    Lead use assemblies (LUAs) of high burnup 8x8 fuel design for Japanese BWRs were irradiated up to 5 cycles in Fukushima Daini Nuclear Power Station No. 2 Unit. Segment rods were installed in LUAs and used for power ramp tests in Japanese Material Test Reactor (JMTR). Post irradiation examinations (PIEs) of segment rods were carried out at Nippon Nuclear Fuel Development Co., Ltd. before and after ramp tests. Maximum linear heat rates of LUAs were kept above 300 W/cm in the first cycle, above 250 W/cm in the second and third cycles and decreased to 200 W/cm in the fourth cycle and 80 W/cm in the fifth cycle. The integrity of high burnup 8x8 fuel was confirmed up to the bundle burnup of 48 GWd/t after 5 cycles of irradiation. Systematic and high quality data were collected through detailed PIEs. The main results are as follows. The oxide on the outer surface of cladding tubes was uniform and its thickness was less than 20 micro-meter after 5 cycles of irradiation and was almost independent of burnup. Hydrogen contents in cladding tubes were less than 150 ppm after 5 cycles of irradiation, although hydrogen contents increased during the fourth and fifth irradiation cycles. Mechanical properties of cladding tubes were on the extrapolated line of previous data up to 5 cycles of irradiation. Fission gas release rates were in the low level (mainly less than 6%) up to 5 cycles of irradiation due to the design to decrease pellet temperature. Pellet-cladding bonding layers were observed after the third cycle and almost full bonding was observed after the fifth cycle. Pellet volume increased with burnup in proportion to solid swelling rate up to the forth cycle. After the fifth cycle, slightly higher pellet swelling was confirmed. Power ramp tests were carried out and satisfactory performance of Zr-lined cladding tube was confirmed up to 60 GWd/t (segment average burnup). One segment rod irradiated for 3 cycles failed by a single step ramp test at terminal ramp power of 614 W

  11. Axial myopathy

    DEFF Research Database (Denmark)

    Witting, Nanna; Andersen, Linda K; Vissing, John

    2016-01-01

    Classically, myopathies are categorized according to limb or cranial nerve muscle affection, but with the growing use of magnetic resonance imaging it has become evident that many well-known myopathies have significant involvement of the axial musculature. New disease entities with selective axial...

  12. High burnup MOX fuel assembly

    International Nuclear Information System (INIS)

    Blanpain, P.; Brunel, L.

    1999-01-01

    From the outset, the MOX product was required to have the same performance as UO 2 in terms of burnup and operational flexibility. In fact during the first years the UO 2 managements could not be applied to MOX. The changeover to an AFA 2G type fuel allowed an improvement in NPP operational flexibility. The move to the AFA 3G design fuel will enable an increase in the burnup of the MOX assemblies to the level of the UO 2 ones ('MOX Parity' project). But the FRAMATOME fuel development objective does not stop at the obtaining of parity between the current MOX and UO 2 products: this parity must remain guaranteed and the MOX managements must evolve in the same way as the UO 2 managements. The goal of the MOX product development programmes underway with COGEMA and the CEA is the demonstration over the next 10 years of a fuel capable of reaching burnups of 70 GWD/T. The research programmes focus on the fission gas release aspect, with three issues explored: optimization of pellet microstructures and validation in experimental reactor ; build-up of experience feedback from fission gas release at elevated burnups in commercial reactors, both for current and experimental products; adaptation and qualification of the design models and tools, over the ranges and for the products concerned. The product arising from these development programmes should be offered on the market around 2010. While meeting safety requirements, it will cater for the needs of the utilities in terms of product reliability, personnel dosimetry and kWh output costs (increase in burnup, NPP maneuverability and availability, minimization of process waste). (authors)

  13. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning

    International Nuclear Information System (INIS)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-01-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  14. Observations on the CANDLE burn-up in various geometries

    International Nuclear Information System (INIS)

    Seifritz, W.

    2007-01-01

    We have looked at all geometrical conditions under which an auto catalytically propagating burnup wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning 'spaghetti' reactor and the azimuthally burning 'pancake' reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometry, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed. A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology

  15. Effect of burnup history by moderator density on neutron-physical characteristics of WWER-1000 core

    International Nuclear Information System (INIS)

    Ovdiienko, I.; Kuchin, A.; Khalimonchuk, V.; Ieremenko, M.

    2011-01-01

    Results of assessment of burnup history effect by moderator density on neutron physical characteristics of WWER-1000 core are presented on example of stationary fuel loading with Russian design fuel assembly TWSA and AER benchmark for Khmelnitsky NPP that was proposed by TUV and SSTC NRC at nineteenth symposium. Assessment was performed by DYN3D code and cross section library sets generated by HELIOS code. Burnup history was taken into account by preparing of numerous cross section sets with different isotopic composition each of which was obtained by burning under different moderator density. For analysis of history effect 20 cross section sets were prepared for each fuel assembly corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect. Main attention of study was paid to effect of burnup history by moderator density to axial power distribution. Results of study for AER benchmark were compared with experimental values of axial power distribution for fuel assemblies of first, second, third and fourth year operation. (Authors)

  16. Assessment of axial gamma dose rate profile on irradiated fuel assembly using polycarbonate film and perspex dosimeters

    International Nuclear Information System (INIS)

    Joshi, V.B.; Janardhanan, S.; Pillai, P.R.; Somanathan, K.; Narayan, K.K.; John, J.; Kutty, K.N.; Deo, V.R.; Popli, O.L.

    1986-01-01

    The dose-rate profile of irradiated fuel rod is required for optimisation of radiation shielding from safety point of view during storage, handling and metallurgical examination. Since the dose-rates are in kilogray per hour, their determination requires special evaluation techniques. This paper illustrates the application of Makrofol-N and red perspex (AERE 4034B) for this purpose. They are compared with CaSO 4 :Dy thermoluminescence dosimeter. (author). 4 refs

  17. Burnup credit activities in the United States

    International Nuclear Information System (INIS)

    Lake, W.H.; Thomas, D.A.; Doering, T.W.

    2001-01-01

    This report covers progress in burnup credit activities that have occurred in the United States of America (USA) since the International Atomic Energy Agency's (IAEA's) Advisory Group Meeting (AGM) on Burnup Credit was convened in October 1997. The Proceeding of the AGM were issued in April 1998 (IAEA-TECDOC-1013, April 1998). The three applications of the use of burnup credit that are discussed in this report are spent fuel storage, spent fuel transportation, and spent fuel disposal. (author)

  18. PENBURN - A 3-D Zone-Based Depletion/Burnup Solver

    International Nuclear Information System (INIS)

    Manalo, Kevin; Plower, Thomas; Rowe, Mireille; Mock, Travis; Sjoden, Glenn E.

    2008-01-01

    PENBURN (Parallel Environment Burnup) is a general depletion/burnup solver which, when provided with zone-based reaction rates, computes time-dependent isotope concentrations for a set of actinides and fission products. Burnup analysis in PENBURN is performed with a direct Bateman-solver chain solution technique. Specifically, in tandem with PENBURN is the use of PENTRAN, a parallel multi-group anisotropic Sn code for 3-D Cartesian geometries. In PENBURN, the linear chain method is actively used to solve individual isotope chains which are then fully attributed by the burnup code to yield integrated isotope concentrations for each nuclide specified. Included with the discussion of code features, a single PWR fuel pin calculation with the burnup code is performed and detailed with a benchmark comparison to PIE (Post-Irradiation Examination) data within the SFCOMPO (Spent Fuel Composition / NEA) database, and also with burnup codes in SCALE5.1. Conclusions within the paper detail, in PENBURN, the accuracy of major actinides, flux profile behavior as a function of burnup, and criticality calculations for the PWR fuel pin model. (authors)

  19. Application of burnup credit concept to transport

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Nakagome, Yoshihiro.

    1994-01-01

    For the design and safety assessment of the casks for transporting spent fuel, the fuel contained in them has been assumed to be new fuel. The reason is, it was difficult to evaluate the variation of the reactivity of fuel, and the research on the affecting factors and the method of measuring burnup were not much advanced. Recently, high burnup fuel has been adopted, and initial degree of enrichment rose. The research has been advanced for pursuing the economy of the casks for spent fuel, and burnup credit has become applicable to their design and safety assessment. As the result, the containing capacity increases by about 20%. When burnup credit is considered, it is necessary to confirm accurately the burnup of spent fuel. The burnup dependence of the concentration of fissile substances and neutron emissivity, the coolant void dependence of the concentration of fissile substances, and the relation of neutron multiplication rate with initial degree of enrichment or burnup are discussed. The conceptual design of casks considering burnup credit and its assessment, the merit, problem and the countermeasures to it when burnup credit is introduced are described. (K.I.)

  20. Phenomena and parameters important to burnup credit

    International Nuclear Information System (INIS)

    Parks, C.V.; Dehart, M.D.; Wagner, J.C.

    2001-01-01

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water- reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the United States and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given. (author)

  1. Axial velocity profiles and secondary flows of developing laminar flows in a straight connected exit region of a 180 .deg. square curved duct

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Hyun Chull; Lee, Heang Nam; Park, Gil Moon [Chosun Univ., Gwangju (Korea, Republic of)

    2005-10-01

    In the present study, characteristics of steady state laminar flows of a straight duct connected to a 180 .deg. curved duct were examined in the entrance region through experimental and numerical analyses. For the analysis, the governing equations of laminar flows in the Cartesian coordinate system were applied. Flow characteristics such as velocity profiles and secondary flows were investigated numerically and experimentally in a square cross-sectional straight duct by the PIV system and a CFD code (STARCD). For the PIV measurement, smoke particles produced from mosquito coils. The experimental data were obtained at 9 points dividing the test sections by 400 mm. Experimental and numerical results can be summarized as follows. 1) Reynolds number, Re was increased, dimensionless velocity profiles at the outer wall were increased due to the effect of the centrifugal force and secondary flows. 2) The intensity of a secondary flow became stronger at the inner wall rather than the outer wall regardless of Reynolds number. Especially, fluid dynamic phenomenon called conner impact were observed at dimensionless axial position, x/D{sub h}=50.

  2. Axial heterogeneous core concept applied for super phoenix reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

    1991-11-01

    Always maintaining the current design rules, this paper presents a parametric study on the type of axial heterogeneous core concept (CHA), utilizing a core of fast reactor Super Phenix type, reaching a maximum thermal burnup rate of 150000 M W d/t and being managed in single batch. (author)

  3. Experimental and Numerical Study of the Buckling of Composite Profiles with Open Cross Section under Axial Compression

    Science.gov (United States)

    Rozylo, Patryk; Teter, Andrzej; Debski, Hubert; Wysmulski, Pawel; Falkowicz, Katarzyna

    2017-10-01

    The object of the research are short, thin-walled columns with an open top-hat cross section made of multilayer laminate. The walls of the investigated profiles are made of plate elements. The entire columns are subjected to uniform compression. A detailed analysis allowed us to determine critical forces and post-critical equilibrium paths. It is assumed that the columns are articulately supported on the edges forming their ends. The numerical investigation is performed by the finite element method. The study involves solving the problem of eigenvalue and the non-linear problem of stability of the structure. The numerical analysis is performed by the commercial simulation software ABAQUS®. The numerical results are then validated experimentally. In the discussed cases, it is assumed that the material operates within a linearly-elastic range, and the non-linearity of the FEM model is due to large displacements.

  4. Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

    International Nuclear Information System (INIS)

    Mizuno, M.; Enokido, Y.; Itaki, T.; Kono, K.; Unno, I.; Yamanouchi, S.

    1985-01-01

    A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis

  5. Study on small long-life LBE cooled fast reactor with CANDLE burn-up. Part 1. Steady state research

    International Nuclear Information System (INIS)

    Yan, Mingyu; Sekimoto, Hiroshi

    2008-01-01

    Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is used as coolant. From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%. (author)

  6. High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

    International Nuclear Information System (INIS)

    Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

    2013-01-01

    In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. Burnup determination of mass spectrometry for nuclear fuels

    International Nuclear Information System (INIS)

    Zhang Chunhua.

    1987-01-01

    The various methods currently being used in burnup determination of nuclear fuels are studied and reviewed. The mass spectrometry method of destructive testing is discussed emphatically. The burnup determination of mass spectrometry includes heavy isotopic abundance ratio method and isotope dilution mass spectrometry used as burnup indicator for the fission products. The former is applied to high burnup level, but the later to various burnup level. According to experiences, some problems which should be noticed in burnup determination of mass spectrometry are presented

  10. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-01

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

  11. Issues for effective implementation of burnup credit

    International Nuclear Information System (INIS)

    Parks, C.V.; Wagner, J.C.

    2001-01-01

    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. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

    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

  13. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  14. Regulatory status of burnup credit for dry storage and transport of spent nuclear fuel in the United States

    International Nuclear Information System (INIS)

    Carlson, D.E.

    2001-01-01

    During 1999, the Spent Fuel Project Office of the U.S. Nuclear Regulatory Commission (NRC) introduced technical guidance for allowing burnup credit in the criticality safety analysis of casks for transporting or storing spent fuel from pressurized water reactors. This paper presents the recommendations embodied by the current NRC guidance, discusses associated technical issues, and reviews information needs and industry priorities for expanding the scope and content of the guidance. Allowable analysis approaches for burnup credit must account for the fuel irradiation variables that affect spent fuel reactivity, including the axial and horizontal variation of burnup within fuel assemblies. Consistent with international transport regulations, the burnup of each fuel assembly must be verified by pre-loading measurements. The current guidance limits the credited burnup to no more than 40 GWd/MTU and the credited cooling time to five years, imposes a burnup offset for fuels with initial enrichments between 4 and 5 wt% 235U, does not include credit for fission products, and excludes burnup credit for damaged fuels and fuels that have used burnable absorbers. Burnup credit outside these limits may be considered when adequately supported by technical information beyond that reviewed to-date by the NRC staff. The guidance further recommends that residual subcritical margins from the neglect of fission products, and any other nuclides not credited in the licensing-basis analysis, be estimated for each cask design and compared against estimates of the maximum reactivity effects associated with remaining computational uncertainties and potentially nonconservative modeling assumptions. The NRC's Office of Nuclear Regulatory Research is conducting a research program to help develop the technical information needed for refining and expanding the evolving guidance. Cask vendors have announced plans to submit the first NRC license applications for burnup credit later this year

  15. First steps towards modelling high burnup effect in UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    O` Carroll, C; Lassmann, K; Laar, J Van De; Walker, C T [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    High burnup initiates a process that can lead to major microstructural changes near the edge of the fuel: formation of subgrains, the loss of matrix fission gas and an increase in porosity. A consequence of this, is a decrease of thermal conductivity near the edge of the fuel which may be major implications for the performance of LWR fuels at higher burnup. The mechanism for the changes in grain structure, the apparent depletion of Xe and increase in porosity is associated with the high fission density at the fuel periphery. This is in turn due to the preferential capture of epithermal neutrons in the resonances of {sup 238}U. The new model TUBRNP predicts the radial burnup profile as a function of time together with the radial profile of plutonium. The model has been validated with data from LWR UO{sub 2} fuels with enrichments in the range 2 to 8.25% and burnups between 21 to 75 Gwd/t. It has been reported that at high burnup EPMA measures a sharp decrease in the concentration of Xe near the fuel surface. This loss of Xe is interpreted as a signal that the gas has been swept out of the original grains into pores: this ``missing`` Xe has been measured by XRF. It has been noted experimentally that the restructuring (Xe depletion and changes in grain structure) have an onset threshold local burnup in the region of 70 to 80 GWd/t: a specific value was taken for use in the model. For a given fuel TUBRNP predicts the local burnup profile, and the depth corresponding to the threshold value is taken to be the thickness of the Xe depleted region. The theoretical predictions have been compared with experimental data. The results are presented and should be seen as a first step in the development of a more detailed model of this phenomenon. (author). 22 refs, 9 figs, 2 tabs.

  16. Value of burnup credit beyond actinides

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.; Kang, Chi.

    1997-01-01

    DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs

  17. Burnup verification using the FORK measurement system

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

    Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK measurement system, designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program, has been used to verify reactor site records for burnup and cooling time for many years. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. This report deals with the application of the FORK system to burnup credit operations based on measurements performed on spent fuel assemblies at the Oconee Nuclear Station of Duke Power Company

  18. Threshold burnup for recrystallization and model for rim porosity in the high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong

    1998-01-01

    Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75μm and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup

  19. Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2009-07-01

    The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

  20. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  1. Burnup analysis of the power reactor, 2

    International Nuclear Information System (INIS)

    Ezure, Hideo

    1975-09-01

    In burnup analysis of JPDR-1 with FLARE, it was found to have problems. The program FLORA was developed for solution of the problems. By their bench mark tests FLORA was found to be useful for three-dimensional thermal-hydro-dynamic analysis of BWRs. It was applied to analysis of the burnup of JPDR-1. The input data and option of FLORA were corrected on referring to the results of gammer probe tests for JPDR-1. The void, source and burnup distributions were calculated each month during the operation. The burnup distribution in three assemblies revealed by a destructive test agrees better with that by FLORA than by FLARE. It was shown that the distortion of power distribution around the control rods by FLORA was smaller and closer to that by the gammer probe tests than by FLARE, and the connector of fuel assemblies and the plugs in the reflector had much influence on the power distribution. (auth.)

  2. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

  3. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)

  4. FEMAXI-7 analysis on behavior of medium and high burnup BWR fuels during base-irradiation and power ramp

    Energy Technology Data Exchange (ETDEWEB)

    Ogiyanagi, Jin, E-mail: ohgiyanagi.jin@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. Black-Right-Pointing-Pointer Calculated FGR and cladding deformation showed reasonable agreement with PIE data. Black-Right-Pointing-Pointer High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. Black-Right-Pointing-Pointer Local PCMI model in the code could reasonably predict cladding ridging deformation. - Abstract: Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.

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

    International Nuclear Information System (INIS)

    Wiesenack, W.

    1996-01-01

    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

  6. K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1998-08-01

    This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k inf values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon consultation with the YMP staff, a Quad Cities BWR fuel assembly was selected as a baseline assembly. This design consists of seven axial enrichment zones, three of which contain natural uranium oxide. No attempt was made to find a bounding or even typical assembly design due to the wide variety in fuel assembly designs necessary for consideration. The current work concentrates on establishing a baseline analysis, along with a small number of sensitivity studies which can be expected later if desired. As a result of similar studies of this nature, several effects are known to be important in the determination of the final k inf for spent fuel in a cask-like geometry. For a given enrichment there is an optimal burnup: for lower burnups, excess energy (and corresponding excess reactivity) is present in the fuel assembly; for larger burnups, the assembly is overburned and essentially driven by neighboring fuel assemblies. The majority of the burnup/enrichment scenarios included in this study were for some near-optimum burnup/enrichment combinations as determined from Energy Information Administration (EIA) data. Several calculations were performed for under- and over-burned fuel to show these effects

  7. Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

    International Nuclear Information System (INIS)

    Setiadipura, T.; Zuhair; Irwanto, D.

    2015-01-01

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

  8. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  9. Comparative study on plutonium and MA recycling in equilibrium burnup and standard burnup of PWR

    International Nuclear Information System (INIS)

    Waris, Abdul; Kurniadi, Rizal; Su'ud, Zaki; Permana, Sidik

    2005-01-01

    The equilibrium burnup model is a powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor. However, this method needs to be verified since the method is not a standard tool. The present study aimed to compare the characteristics of plutonium recycling and plutonium and minor actinides (MA) recycling in PWR with the equilibrium burnup and the standard burnup. In order to become more comprehensive study, an influence of moderator-to-fuel volume ratio (MFR) changes by changing the pin-pitch of fuel cell has been evaluated. The MFR ranges from 0.5 to 4.0. For the equilibrium burnup we used equilibrium cell-burnup code. We have employed 1368 nuclides in the equilibrium calculation with 129 of them are heavy metals (HMs). For standard burnup, SRAC2002 code has been utilized with 26 HMs and 66 fission products (FPs). The JENDL 3.2 library has been employed for both burnup schemes. The uranium, plutonium and MA vector, which resulted from the equilibrium burnup are directly used as fuel input composition for the standard burnup calculation. Both burnup results demonstrate that plutonium recycling and plutonium and MA recycling can be conducted safer in tight lattice core. They are also show the similar trend in neutron spectrum, which become harder with the increasing number of recycled heavy nuclides as well as the decreasing of the MFR values. However, there are some discrepancy on the effective multiplication factor and the conversion ratio, especially for the reactor core for MFR ≥ 2.0. (author)

  10. Axial tomography

    International Nuclear Information System (INIS)

    Brueckner, K.A.; Lewis, J.H.

    1979-01-01

    The invention relates to axial tomography, sometimes referred to as cross-sectional x-ray. The apparatus described may utilize the conventional x-ray or ultrasonic source and detector and scanning mechanism for producing the plurality of sets of radiation detector output signals. It has the means for storing the detector output signals in analog form with the signals of one set overlying the signals of another set so that signals resulting from radiation through a zone of the object being examined are summed at a corresponding zone in the storage device, typically an electronic storage tube. The summed signals are read from the storage device with a radially inversely proportional reader producing a second signal for storage, again typically in an electronic storage tube. These signals stored in the second storage device are read with Laplacian relation, with the resultant sigal being a video signal that may be connected to a TV monitor for display of the sectional image. In alternative embodiments, optical film systems and electrostatic systems are utilized. (JTA)

  11. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

  12. Optimization of axial enrichment and gadolinia distributions for BWR fuel under control rod programming, (2)

    International Nuclear Information System (INIS)

    Hida, Kazuki; Yoshioka, Ritsuo

    1992-01-01

    A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance. This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs. (author)

  13. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  14. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  15. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

  16. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning; Perfiles axiales de quemado y fraccion de huecos para calculos de criticidad con credito al quemado para combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-07-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  17. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  18. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  19. Nuclear fuels with high burnup: safety requirements

    International Nuclear Information System (INIS)

    Phuc Tran Dai

    2016-01-01

    Vietnam authorities foresees to build 3 reactors from Russian design (VVER AES 2006) by 2030. In order to prepare the preliminary report on safety analysis the Vietnamese Agency for Radioprotection and Safety has launched an investigation on the behaviour of nuclear fuels at high burnups (up to 60 GWj/tU) that will be those of the new plants. This study deals mainly with the behaviour of the fuel assemblies in case of loss of coolant (LOCA). It appears that for an average burnup of 50 GWj/tU and for the advanced design of the fuel assembly (cladding and materials) safety requirements are fulfilled. For an average burnup of 60 GWj/tU, a list of issues remains to be assessed, among which the impact of clad bursting or the hydrogen embrittlement of the advanced zirconium alloys. (A.C.)

  20. Burnup credit in a dry storage module

    International Nuclear Information System (INIS)

    Thornton, J.R.

    1989-01-01

    Comparison of spent fuel storage expansion options available to Oconee Nuclear Station revealed that dry storage could be economically competitive with transshipment and rod consolidation. Economic competitiveness, however, mandated large unit capacity while existing cask handling facilities at Oconee severely limited size and weight. The dry storage concept determined to best satisfy these conflicting criteria is a 24 pressurized water reactor (PWR) fuel assembly capacity NUTECH Horizontal Modular Storage (NUHOMS) system. The Oconee version of the NUHOMS system takes advantage of burnup credit in demonstrating criticality safety. The burnup credit criticality analysis was performed by Duke Power Company's Design Engineering Department. This paper was prepared to summarize the criticality control design features employed in the Oconee NUHOMS-24P DSC basket and to describe the incentives for pursuing a burnup credit design. Principal criticality design parameters, criteria, and analysis methodology are also presented

  1. Status of burnup credit implementation in Switzerland

    International Nuclear Information System (INIS)

    Grimm, P.

    1998-01-01

    Burnup credit is currently not used for the storage of spent fuel in the reactor pools in Switzerland, but credit is taken for integral burnable absorbers. Interest exists to take credit of burnup in future for the storage in a central away-from-reactor facility presently under construction. For spent fuel transports to foreign reprocessing plants the regulations of the receiving countries must be applied in addition to the Swiss licensing criteria. Burnup credit has been applied by one Swiss PWR utility for such transports in a consistent manner with the licensing practice in the receiving countries. Measurements of reactivity worths of small spent fuel samples in a Swiss zero-power research reactor are at an early stage of planning. (author)

  2. On-line extraction of the variance caused by burn-up in in-core three-dimensional power distribution

    International Nuclear Information System (INIS)

    Wang Yaqi; Luo Zhengpei; Li Fu; Liu Wenfeng

    2001-01-01

    In most of PWRs, the ex-core ion-chambers are the sole real-time sensors to respond to in-core power and its axial offset. However, the calibration coefficient of the ion-chambers depends on the (3D) power distribution and varies with the burn-up. People expect to know the variance in distribution caused by burn-up directly from the signals of ion-chambers. This expectation is not realized as yet, because an ion-chamber almost only responds to its nearest fuel assemblies. The authors then developed a two-step method for burn-up characteristic extraction: the harmonics synthesis method and harmonics' burn-up grouping. Using the extracted burn-up characteristics, the relationship between the readings of the ex-core ion-chambers and the in-core 3D power distribution is set up. Through the simulation on the heating reactor, the method of burn-up characteristic extraction is verified under engineering conditions. It is possible to on-line extract the variance caused by burn-up in 3D power distribution

  3. Advancements in reactor physics modelling methodology of Monte Carlo Burnup Code MCB dedicated to higher simulation fidelity of HTR cores

    International Nuclear Information System (INIS)

    Cetnar, Jerzy

    2014-01-01

    The recent development of MCB - Monte Carlo Continuous Energy Burn-up code is directed towards advanced description of modern reactors, including double heterogeneity structures that exist in HTR-s. In this, we exploit the advantages of MCB methodology in integrated approach, where physics, neutronics, burnup, reprocessing, non-stationary process modeling (control rod operation) and refined spatial modeling are carried in a single flow. This approach allows for implementations of advanced statistical options like analysis of error propagation, perturbation in time domain, sensitivity and source convergence analyses. It includes statistical analysis of burnup process, emitted particle collection, thermal-hydraulic coupling, automatic power profile calculations, advanced procedures of burnup step normalization and enhanced post processing capabilities. (author)

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

    International Nuclear Information System (INIS)

    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

  5. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  6. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    Matausek, M.

    1984-01-01

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

  7. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  8. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  9. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  10. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

  11. Burnup verification tests with the FORK measurement system-implementation for burnup credit

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

    Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. It was designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program and is well suited to verify burnup and cooling time records at commercial Pressurized Water Reactor (PWR) sites. This report deals with the application of the FORK system to burnup credit operations

  12. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    International Nuclear Information System (INIS)

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-01-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  13. Fuel analysis code FAIR and its high burnup modelling capabilities

    International Nuclear Information System (INIS)

    Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1995-01-01

    A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

  14. Calculation of axial hydrogen redistribution on the spent fuels during interim dry storage

    International Nuclear Information System (INIS)

    Sasahara, Akihiro; Matsumura, Tetsuo

    2006-01-01

    One of the phenomena that will affect fuel integrity during a spent fuel dry storage is a hydrogen axial migration in cladding. If there is a hydrogen pickup in cladding in reactor operation, hydrogen will move from hotter to colder cladding region in the axial direction under fuel temperature gradient during dry storage. Then hydrogen beyond solubility limit in colder region will be precipitated as hydride, and consequently hydride embrittlement may take place in the cladding. In this study, hydrogen redistribution experiments were carried out to obtain the data related to hydrogen axial migration by using actually twenty years dry (air) stored spent PWR-UO 2 fuel rods of which burn-ups were 31 and 58 MWd/kg HM. From the hydrogen redistribution experiments, the heat of transport of hydrogen of zircaloy-4 cladding from twenty years dry stored spent PWR-UO 2 fuel rods were from 10.1 to 18.6 kcal/mol and they were significantly larger than that of unirradiated zircaloy-4 cladding. This means that hydrogen in irradiated cladding can move easier than that in unirradiated cladding. In the hydrogen redistribution experiments, hydrogen diffusion coefficients and solubility limit were also obtained. There are few differences in the diffusion coefficients and solubility limits between the irradiated cladding and unirradiated cladding. The hydrogen redistribution in the cladding after dry storage for forty years was evaluated by one-dimensional diffusion calculation using the measured values. The maximum values as the heat of transports, diffusion coefficients and solubility limits of the irradiated cladding and various spent fuel temperature profiles reported were used in the calculation. The axial hydrogen migration was not significant after dry storage for forty years in helium atmosphere and the maximum values as the heat of transports, diffusion coefficients and solubility limits of the unirradiated cladding gave conservative evaluation for hydrogen redistribution

  15. Burnup calculation for a tokamak commercial hybrid reactor

    International Nuclear Information System (INIS)

    Feng Kaiming; Xie Zhongyou

    1990-08-01

    A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

  16. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

    Lyric, Zoairia Idris; Mahmood, Mohammad Sayem; Motalab, Mohammad Abdul; Khan, Jahirul Haque

    2013-01-01

    Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor

  17. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  18. Burnup credit effect on proposed cask payloads

    International Nuclear Information System (INIS)

    Hall, I.K.

    1989-01-01

    The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems which will allow safe and economical movement of commercial spent nuclear fuel and high-level waste from the generator to the Federal repository or Monitored Retrievable Storage (MRS) facility. Program schedule objectives for the initial phase of the CSDP include the development of certified spent fuel cask systems by 1995 to support Office of Civilian Radioactive Waste Management shipments from the utilities beginning in the late 1990s. Forty-nine proposals for developing a family of spent fuel casks were received and comparisons made. General conclusions that can be drawn from the comparisons are that (1) the new generation of casks will have substantially increased payloads in comparison to current casks, and (2) an even greater payload increase may be achievable with burnup credit. The ranges in the payload estimates do not allow a precise separation of the payload increase attributable to the proposed allowance of fuel burnup credit, as compared wilt the no-burnup-credit case. The beneficial effects of cask payload increases on overall costs and risks of transporting spent fuel are significant; therefore further work aimed toward taking advantage of burnup credit is warranted

  19. WWER-1000 Burnup Credit Benchmark (CB5)

    International Nuclear Information System (INIS)

    Manolova, M.A.

    2002-01-01

    In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)

  20. Whole core burnup calculations using 'MCNP'

    International Nuclear Information System (INIS)

    Haran, O.; Shaham, Y.

    1996-01-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

  1. Whole core burnup calculations using `MCNP`

    Energy Technology Data Exchange (ETDEWEB)

    Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

  2. Burnup calculations using Monte Carlo method

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Degweker, S.B.

    2009-01-01

    In the recent years, interest in burnup calculations using Monte Carlo methods has gained momentum. Previous burn up codes have used multigroup transport theory based calculations followed by diffusion theory based core calculations for the neutronic portion of codes. The transport theory methods invariably make approximations with regard to treatment of the energy and angle variables involved in scattering, besides approximations related to geometry simplification. Cell homogenisation to produce diffusion, theory parameters adds to these approximations. Moreover, while diffusion theory works for most reactors, it does not produce accurate results in systems that have strong gradients, strong absorbers or large voids. Also, diffusion theory codes are geometry limited (rectangular, hexagonal, cylindrical, and spherical coordinates). Monte Carlo methods are ideal to solve very heterogeneous reactors and/or lattices/assemblies in which considerable burnable poisons are used. The key feature of this approach is that Monte Carlo methods permit essentially 'exact' modeling of all geometrical detail, without resort to ene and spatial homogenization of neutron cross sections. Monte Carlo method would also be better for in Accelerator Driven Systems (ADS) which could have strong gradients due to the external source and a sub-critical assembly. To meet the demand for an accurate burnup code, we have developed a Monte Carlo burnup calculation code system in which Monte Carlo neutron transport code is coupled with a versatile code (McBurn) for calculating the buildup and decay of nuclides in nuclear materials. McBurn is developed from scratch by the authors. In this article we will discuss our effort in developing the continuous energy Monte Carlo burn-up code, McBurn. McBurn is intended for entire reactor core as well as for unit cells and assemblies. Generally, McBurn can do burnup of any geometrical system which can be handled by the underlying Monte Carlo transport code

  3. MTR core loading pattern optimization using burnup dependent group constants

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

    Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

  4. Influence of prolonged nuclear fuel burnup on safety characteristics of advanced PWRs

    International Nuclear Information System (INIS)

    Spasojevic, D.; Matausek, M.; Marinkovic, N.

    1989-01-01

    Prolonged nuclear fuel burnup in advanced NPP with four or more instead of three one-year cycles, and/or with 15- to 18-month instead of standard 12-month cycles, requires the fresh fuel to have increased enrichment combined with burnable poisons. This causes changes in axial and radial distribution of power generation during the particular fuel cycles, so that detailed analyses of thermal reliability of reactor core becomes necessary. This paper presents the results of the analysis of the departure from nuclear boiling ratio DNBR for an equilibrium cycle of an advanced PWR. (author)

  5. Measuring device for the distribution of burn-up degree in fuel assembly irradiated in nuclear reactor

    International Nuclear Information System (INIS)

    Kumanomido, Hironori

    1989-01-01

    The object of the invention is to measure the distribution of burn-up degree, of fuel assemblies irradiated in a nuclear reactor in a short time and exactly. That is, the device comprises a device main body having substantially the same length as that for the axial length of a fuel assembly and a detector container disposed axially slidably to the main body. A plurality of radiation detectors are arranged at an equi-axial pitch and contained in the container. The container is caused to slide at a pitch equal to the equi-axial distance of the detectors. In the device having thus been constituted, measurement is conducted at least for twice at an axial position on the side of a fuel assembly irradiated in the nuclear reactor and a position caused to slide therefrom by one pitch. Based on the result, the sensitivities between each of the detectors are compared and the relative sensitivity of the radiation detectors is calibrated. Accordingly, the sensitivity between each of the detectors can be calibrated rapidly and easily. As a result, the distribution of the burn-up degree, etc of irradiated fuel assembly can be measured exactly. (K.M.)

  6. Burnup credit activities being conducted in the United States

    International Nuclear Information System (INIS)

    Lake, W.

    1998-01-01

    The paper describes burnup credit activities being conducted in the U.S. where burnup credit is either being used or being planned to be used for storage, transport, and disposal of spent nuclear fuel. Currently approved uses of burnup credit are for wet storage of PWR fuel. For dry storage of spent PWR fuel, burnup credit is used to supplement a principle of moderator exclusion. These storage applications have been pursued by the private sector. The Department of Energy (DOE) which is an organization of the U.S. Federal government is seeking approval for burnup credit for transport and disposal applications. For transport of spent fuel, regulatory review of an actinide-only PWR burnup credit method is now being conducted. A request by DOE for regulatory review of actinide and fission product burnup credit for disposal of spent BWR and PWR fuel is scheduled to occur in 1998. (author)

  7. Core burn-up calculation method of JRR-3

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Yamashita, Kiyonobu

    2007-01-01

    SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)

  8. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    Tran, Viet-Phu; Tran, Hoai-Nam; Yamamoto, Akio; Endo, Tomohiro

    2017-01-01

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO_2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

  9. Automated generation of burnup chain for reactor analysis applications

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Viet-Phu [VINATOM, Hanoi (Viet Nam). Inst. for Nuclear Science and Technology; Tran, Hoai-Nam [Duy Tan Univ., Da Nang (Viet Nam). Inst. of Research and Development; Yamamoto, Akio; Endo, Tomohiro [Nagoya Univ., Nagoya-shi (Japan). Dept. of Materials, Physics and Energy Engineering

    2017-05-15

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO{sub 2} and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

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

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    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)

  11. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    Finck, P.J.; Stenberg, C.G.

    1998-01-01

    The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work

  12. Resolution of axial shear strain elastography

    International Nuclear Information System (INIS)

    Thitaikumar, Arun; Righetti, Raffaella; Krouskop, Thomas A; Ophir, Jonathan

    2006-01-01

    The technique of mapping the local axial component of the shear strain due to quasi-static axial compression is defined as axial shear strain elastography. In this paper, the spatial resolution of axial shear strain elastography is investigated through simulations, using an elastically stiff cylindrical lesion embedded in a homogeneously softer background. Resolution was defined as the smallest size of the inclusion for which the strain value at the inclusion/background interface was greater than the average of the axial shear strain values at the interface and inside the inclusion. The resolution was measured from the axial shear strain profile oriented at 45 0 to the axis of beam propagation, due to the absence of axial shear strain along the normal directions. The effects of the ultrasound system parameters such as bandwidth, beamwidth and transducer element pitch along with signal processing parameters such as correlation window length (W) and axial shift (ΔW) on the estimated resolution were investigated. The results show that the resolution (at 45 0 orientation) is determined by the bandwidth and the beamwidth. However, the upper bound on the resolution is limited by the larger of the beamwidth and the window length, which is scaled inversely to the bandwidth. The results also show that the resolution is proportional to the pitch and not significantly affected by the axial window shift

  13. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Pavelescu, M.; Borza, M.

    1975-01-01

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

  14. Conceptual cask design with burnup credit

    International Nuclear Information System (INIS)

    Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong

    2003-01-01

    Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)

  15. Time step length versus efficiency of Monte Carlo burnup calculations

    International Nuclear Information System (INIS)

    Dufek, Jan; Valtavirta, Ville

    2014-01-01

    Highlights: • Time step length largely affects efficiency of MC burnup calculations. • Efficiency of MC burnup calculations improves with decreasing time step length. • Results were obtained from SIE-based Monte Carlo burnup calculations. - Abstract: We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy

  16. A Monte Carlo burnup code linking MCNP and REBUS

    International Nuclear Information System (INIS)

    Hanan, N. A.

    1998-01-01

    The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented

  17. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1993-01-01

    The use of burnup credit in the criticality safety analysis of the GA-4 Cask increases the cask's capacity from three spent fuel assemblies to four, resulting in reduced public and occupational risk and reduced life cycle costs. GA's criticality calculations for burnup credit, including the associated uncertainties and analytical bias, establish the minimum burnup required as a function of initial enrichment to maintain K eff ≤ 0.95 under any conceivable condition. The minimum burnup requirement as a function of initial enrichment has been determined to be 15,000 MWd/MTU for 3.5 wt% U-235 fuel, 20,000 MWd/MTU for 4.0 wt% U-235 fuel and 25,000 MWd/MTU for 4.5 wt% U-235 fuel. The minimum burnup requirement as a function of enrichment is well below the typical burnup levels seen in the current and projected spent fuel inventory. (J.P.N.)

  18. Implementation of burnup credit in spent fuel management systems

    International Nuclear Information System (INIS)

    Dyck, H.P.

    2001-01-01

    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)

  19. Theory analysis and simple calculation of travelling wave burnup scheme

    International Nuclear Information System (INIS)

    Zhang Jian; Yu Hong; Gang Zhi

    2012-01-01

    Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)

  20. Burnup calculation in microcells of high conversion reactors

    International Nuclear Information System (INIS)

    Gomez, S.E.; Salvatore, M.; Patino, N.E.; Abbate, M.J.

    1991-01-01

    The development of high converter reactors (HCR) requires careful burnup calculations because their main goals are reach high discharge burnup levels (Up to 50 GWd/T) and a close to one conversion ratio. Then, it is necessary a revision of design elements used for this type of calculation. In this work, a burnup module (BUM) developed in order to use nuclear data directly from evaluated data files is presented; these was included in the AMPX system. (author)

  1. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

    1990-12-01

    A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

  2. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

    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

  3. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1983-01-01

    Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt

  4. Determination of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Kristak, J.; Vobecky, M.

    1973-01-01

    Samples containing a known content of 235 U were irradiated with several different neutron doses and activities were determined of radionuclides including 125 Sb, 144 Ce, 134 Cs, 154 Eu, 103 Ru, 95 Zr. The values thus obtained were divided by the 137 Cs activity value. The resulting neutron dose-dependent value is plotted into a calibration graph. The degree of nuclear fuel burn-up is obtained from the graph using an experimentally determined ratio of the activities of the above radionuclides. (B.S.)

  5. 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.

  6. Technological and licensing challenges for high burnup fuel

    International Nuclear Information System (INIS)

    Gross, H.; Urban, P.; Fenzlein, C.

    2002-01-01

    Deregulation of electricity markets is driving electricity prices downward as well in the U.S. as in Europe. As a consequence high burnup fuel will be demanded by utilities using either the storage or the reprocessing option. At a minimum, burnups consistent with the current political enrichment limit of 5 w/o will be required for both markets.Significant progress has been achieved in the past by Siemens in meeting the demands of utilities for increased fuel burnup. The technological challenges posed by the increased burnup are mainly related to the corrosion and hydrogen pickup of the clad, the high burnup properties of the fuel and the dimensional changes of the fuel assembly structure. Clad materials with increased corrosion resistance appropriate for high burnup have been developed. The high burnup behaviour of the fuel has been extensively investigated and the decrease of thermal conductivity with burnup, the rim effect of the pellet and the increase of fission gas release with burnup can be described, with good accuracy, in fuel rod computer codes. Advanced statistical design methods have been developed and introduced. Materials with increased corrosion resistance are also helpful controlling the dimensional changes of the fuel assembly structure. In summary, most of the questions about the fuel operational behaviour and reliability in the high burnup range have been solved - some of them are still in the process of verification - or the solutions are visible. This fact is largely acknowledged by regulators too. The main licensing challenges for high burnup fuel are currently seen for accident condition analyses, especially for RIA and LOCA. (author)

  7. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    International Nuclear Information System (INIS)

    Chung, H. M.

    2000-01-01

    Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10 21 n cm -2 to 5.9 x 10 21 n cm -2 (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed

  8. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

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

  9. Establishing a PWR burn-up library

    International Nuclear Information System (INIS)

    Lutz, D.C.

    1981-01-01

    Starting out from data file ENDF/B IV /1/, a cross-section library has been established for the calculation of operating conditions in pressurized water reactors of the type used in BIBLIS B. The library includes macroscopic, homogenized 2-group cross-sections for all types of fuel elements used in this reactor, including those equipped with boron glass rods. For their calculation the previous irradiation of the fuel has been taken into consideration by approximation. Information on fuel consumption from cell burn-up calculations has been stored in a separate data file. It was designed as a base for the determination of cross sections to be used in the calculation of the incident ''main-steam pipe fracture''. For this library the description of cross sections as a function of the moderator status chose the water densities at 300 0 C/155 bar, 190 0 C/140 bar and 100 0 C/100 bar as fixed values. The burn-up library has been tested by a three-dimensional calculation for the 1sup(st) cycle of the BIBLIS B-reactor using program QUABOX /2/. This showed variances with the anticipated course concerning critically, which can be explained almost quantitatively by known deficiencies of the ENDF/b-IV library. (orig.) [de

  10. Device for measuring a burnup degree

    International Nuclear Information System (INIS)

    Ito, Toshiaki; Goto, Seiichiro

    1979-01-01

    Purpose: To measure the burnup degree at high efficiency and accuracy. Constitution: The outer metal wall of fuel assemblies is heated under gamma radiation with long half life gamma rays in inverse proportion to the burnup degree and issues infrared radiation in proportion to the intensity of the gamma rays. An image pick-up tube is opposed to one surface of the fuel assemblies to detect the radiated infrared rays. Since the output signal from the pick-up tube is subjected to the absorptive damping by the distance between the pick-up tube and the fuel assembly, as well as water filled in the gap therebetween, it is corrected through a main amplifier comprising a signal correction circuit composed of a characteristic section inverse to the absorption property and a characteristic section inverse to the square of the distance. The corrected output signal is displayed on a display unit such as CRT or recorded in a film or a magnetic tape. (Furukawa, Y.)

  11. Actinides burnup in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

  12. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

  13. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  14. Benchmarking burnup reconstruction methods for dynamically operated research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sternat, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Charlton, William S. [Univ. of Nebraska, Lincoln, NE (United States). National Strategic Research Institute; Nichols, Theodore F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-01

    The burnup of an HEU fueled dynamically operated research reactor, the Oak Ridge Research Reactor, was experimentally reconstructed using two different analytic methodologies and a suite of signature isotopes to evaluate techniques for estimating burnup for research reactor fuel. The methods studied include using individual signature isotopes and the complete mass spectrometry spectrum to recover the sample’s burnup. The individual, or sets of, isotopes include 148Nd, 137Cs+137Ba, 139La, and 145Nd+146Nd. The storage documentation from the analyzed fuel material provided two different measures of burnup: burnup percentage and the total power generated from the assembly in MWd. When normalized to conventional units, these two references differed by 7.8% (395.42GWd/MTHM and 426.27GWd/MTHM) in the resulting burnup for the spent fuel element used in the benchmark. Among all methods being evaluated, the results were within 11.3% of either reference burnup. The results were mixed in closeness to both reference burnups; however, consistent results were achieved from all three experimental samples.

  15. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  16. Burn-up measurements coupling gamma spectrometry and neutron measurement

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H.; Pin, P. [AREVA/CANBERRA, 1 rue des Herons, 78182 St Quentin-en-Yvelines Cedex (France); Lebrun, A. [IAEA, Wagramer Strasse 5, PO Box 100, Vienna (Austria); Oriol, L.; Saurel, N. [CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Gain, T. [AREVA/COGEMA Reprocessing Business Unit, La Hague, 50444 Beaumont Hague Cedex (France)

    2006-07-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  17. Burn-up measurements coupling gamma spectrometry and neutron measurement

    International Nuclear Information System (INIS)

    Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.

    2006-01-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  18. REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR

    International Nuclear Information System (INIS)

    Boettcher, W.; Schmidt, E.

    1969-01-01

    cycle costs are evaluated following proposals of the EURATOM Economic Handbook. 3 - Restrictions on the complexity of the problem: Maximum number of groups of channels having rotation symmetry is 60. Maximum Number of groups of channels having specular symmetry is 120. Maximum number of harmonics for the approximation of the axial flux distribution is 19. Highest order of Bessel functions for the approximation of the radial flux distribution is 12. Maximum number of axial pieces of a channel with possibly different neutronic properties is 20. Maximum number of neutron groups: two fast, one thermal. Maximum number of different types of channels in the reactor is 10. Maximum number of burn-up steps characterizing one type of channel is 50

  19. Finite elements for heat diffusion in heterogenous media with axial symmetry-temperature profile in the neighborhood of a hot penetration

    International Nuclear Information System (INIS)

    Garcia, M.J.A.

    1982-07-01

    A computer program was developed in order to solve the Heat condution equation by means the finite element method. The program, named HEAT, was applied to some typical problems which have analytical solutions. The results by this program were further compared to those obtained by other computer programs that also use the finite element method. The main application of the HEAT program was the calculation of temperature profiles in a hot penetration of the contaiment building of Angra I. The results lead us to conclude that temperatures high enough to cause structural weakening of the contaiment concrete will not be reached, which is in accordance with A.C.I. regulations. (Author) [pt

  20. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    Tran Viet Phu; Tran Hoai Nam; Akio Yamamoto; Tomohiro Endo

    2015-01-01

    This paper presents the development of an automated generation of a new burnup chain for reactor analysis applications. The JENDL FP Decay Data File 2011 and Fission Yields Data File 2011 were used as the data sources. The nuclides in the new chain are determined by restrictions of the half-life and cumulative yield of fission products or from a given list. Then, decay modes, branching ratios and fission yields are recalculated taking into account intermediate reactions. The new burnup chain is output according to the format for the SRAC code system. Verification was performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Further development and applications are being planned with the burnup chain code. (author)

  1. Probabilistic assessment of dry transport with burnup credit

    International Nuclear Information System (INIS)

    Lake, W.H.

    2003-01-01

    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)

  2. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-09-01

    General Atomics (GA) has designed two legal weight truck (LWT) casks, the GA-4 and GA-9, to carry four pressurized-water-reactor (PWR) and nine boiling-water-reactor (BWR) fuel assemblies, respectively. GA plans to submit applications for certification to the US Nuclear Regulatory Commission (NRC) for the two casks in mid-1993. GA will include burnup credit analysis in the Safety Analysis Report for Packaging (SARP) for the GA-4 Cask. By including burnup credit in the criticality safety analysis for PWR fuels with initial enrichments above 3% U-235, public and occupation risks are reduced and cost savings are realized. The GA approach to burnup credit analysis incorporates the information produced in the US Department of Energy Burnup Credit Program. This paper describes the application of burnup credit to the criticality control design of the GA-4 Cask

  3. A Monte Carlo burnup code linking MCNP and REBUS

    International Nuclear Information System (INIS)

    Hanan, N.A.; Olson, A.P.; Pond, R.B.; Matos, J.E.

    1998-01-01

    The REBUS-3 burnup code, used in the anl RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented. (author)

  4. Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

  5. Benefits of actinide-only burnup credit for shutdown PWRs

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.; Kang, C.; Rivard, D.

    1998-02-01

    Owners of PWRs that are shutdown prior to resolution of interim storage or permanent disposal issues have to make difficult decisions on what to do with their spent fuel. Maine Yankee is currently evaluating multiple options for spent fuel storage. Their spent fuel pool has 1,434 assemblies. In order to evaluate the value to a utility of actinide-only burnup credit, analysis of the number of canisters required with and without burnup credit was made. In order to perform the analysis, loading curves were developed for the Holtec Hi-Star 100/MPC-32. The MPC-32 is hoped to be representative of future burnup credit designs from many vendors. The loading curves were generated using the actinide-only burnup credit currently under NRC review. The canister was analyzed for full loading (32 assemblies) and with partial loadings of 30 and 28 assemblies. If no burnup credit is used the maximum capacity was assumed to be 24 assemblies. this reduced capacity is due to the space required for flux traps which are needed to sufficiently reduce the canister reactivity for the fresh fuel assumption. Without burnup credit the 1,343 assemblies would require 60 canisters. If all the fuel could be loaded into the 32 assembly canisters only 45 canisters would be required. Although the actinide-only burnup credit approach is very conservative, the total number of canisters required is only 47 which is only two short of the minimum possible number of canisters. The utility is expected to buy the canister and the storage overpack. A reasonable cost estimate for the canister plus overpack is $500,000. Actinide-only burnup credit would save 13 canisters and overpacks which is a savings of about $6.5 million. This savings is somewhat reduced since burnup credit requires a verification measurement of burnup. The measurement costs for these assemblies can be estimated as about $1 million. The net savings would be $5.5 million

  6. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  7. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    Abe, Alfredo Y.; Dos Santos, Adimir

    2002-01-01

    1 - Description of program or function: HAMCIND is a cell burnup code based in a coupling between HAMMER-TECHNION and CINDER. The fission product poisoning is taken into account in an explicit fashion. 2 - Method of solution: The nonlinear coupled set of equations for the neutron transport and nuclide transmutation equations and nuclide transmutation equations in a unit cell is solved by HAMCIND in a quasi-static approach. The spectral transport equation is solved by HAMMER-TECHNION at the beginning of each time-step while the nuclide transmutation equations are solved by CINDER for every time-step. The HAMMER-TECHNION spectral calculations are performed taking into account the fission product contribution to the macroscopic cross sections (fast and thermal), in the inelastic scattering matrix and even in the thermal scattering matrices. 3 - Restrictions on the complexity of the problem: Restrictions and/or limitations for HAMCIND depend upon the local operating system

  8. The commercial impact of burnup increase

    International Nuclear Information System (INIS)

    Fenzlein, C.; Schricker, W.

    2002-01-01

    Deregulation has a dramatic effect on competition in the electricity markets. This will lead to a continued pressure on the prices in virtually all areas of the nuclear fuel cycle and will encourage further optimization, technical and technological progress and innovations with respect to further cost reductions of power production. The permission of direct disposal, in Germany legally granted in 1994 as an alternative to the reprocessing path, made possible cost savings and has consequently resulted in a decline of reprocessing prices. In addition, suppliers as well as operators are making considerable efforts to reduce the disposal costs fraction by optimizing disposal technologies and concepts. The increase of discharge has essentially contributed to the reduction the disposal cost fraction. Compared to former scenarios, the economic potential of burn-up increase is decreasing

  9. Measurement of burnup in FBR MOX fuel irradiated to high burnup

    International Nuclear Information System (INIS)

    Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

    2003-01-01

    The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

  10. Experimental programmes related to high burnup fuel

    International Nuclear Information System (INIS)

    Vasudeva Rao, P.R.; Vidhya, R.; Ananthasivan, K.; Srinivasan, T.G.; Nagarajan, K.

    2002-01-01

    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 LaC 2 , NdC 2 and SmC 2 have been measured by measuring the vapor pressures of CO over the region Ln 2 O 3 -LnC 2 -C during the carbothermic reduction of Ln 2 O 3 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)

  11. High burnup models in computer code fair

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

    1997-08-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

  12. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

    1997-01-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

  13. Computerised Axial Tomography (CAT)

    Science.gov (United States)

    1990-06-01

    Ministry of’ Defence, Defence Research Information Centre, UK. Computerised Axial Tomography ( CAT ) Report Secufty C"uMiauion tide Onadtiicadon (U. R, Cor S...DRIC T 8485 COMPUTERISED AXIAL TOMOGRAPHY ( CAT ) F.P. GENTILE, F. SABETTA, V. TRO1* ISS R 78/4.Rome, 1.5 Mlarch 1978 (from Italian) B Distribution(f...dello Radiazioni ISSN 0390--6477 F.P. GENTILE, F. SABETTA. V. TROI Computerised Axial Tomography ( CAT ) March 15, 1978). This paper is a review of

  14. Development of methods for burn-up calculations for LWR's

    International Nuclear Information System (INIS)

    Jaschik, W.

    1978-01-01

    This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in which the heterogeneous cell structure and self-shielding effects can explicitly be accounted for. Then the cluster burn-up is simulated by adequate correlation of the burn-up data. The advantage of this method is given by - an exact determination of the real spectrum distribution in the individual fuel element clusters; - an exact determination of the burn-up related spectrum variations for each fuel rod and for each burn-up value obtained; - accounting for heterogeneity of the fuel rod cells and the self-shielding in the fuel; high accuracy of the results of a comparably low effort and - simple handling by largely automating the process of computation. Programed realization was achieved by establishing the RSYST modules ABRAJA, MITHOM, and SIMABB and their implementation within the code system. (orig./HP) [de

  15. Highlights on R and D work related to the achievement of high burnup with MOX fuel in commercial reactors

    International Nuclear Information System (INIS)

    Lippens, M.; Maldague, Th.; Basselier, J.; Boulanger, D.; Mertens, L.

    2000-01-01

    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 Pu 239 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 UO 2 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 UO 2 . 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 UO 2 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 UO 2 , 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 UO 2 . 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)

  16. Signatures for axial chromodynamics

    International Nuclear Information System (INIS)

    Pati, J.C.

    1978-07-01

    Within the context of basic left-right symmetry and the hypothesis of unification of weak, electromagnetic and strong forces at a mass level approximately equal to 10 4 -10 6 GeV, relatively light ''mass'' axial gluons, confined or liberated, must be postulated. The authors remark that the existence of such ''light'' axial gluons supplementing the familiar vector octet preserves the successes of QCD, both for deep inelastic processes and charmonium physics. Through the characteristic spin-spin force, generated by their exchange, they may even help resolve some of the discrepancies between vector QCD predictions and charmonium physics. The main remark of this note is that if colour is liberated, not only vector but also axial-vector gluons are produced in high-energy e - e + experiments, e.g. at PETRA and PEP, with fairly large cross-section. Distinctive decay modes of such liberated axial gluons are noted

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

    International Nuclear Information System (INIS)

    Brown, C.; Hesketh, K.W.; Palmer, I.D.

    1998-01-01

    It is clear that in order to maintain competitiveness with UO 2 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 UO 2 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)

  18. A guide to introducing burnup credit, preliminary version (English translation)

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu

    2017-06-01

    There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee. (author)

  19. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    Medun, V.

    2001-01-01

    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)

  20. Burnup calculation methodology in the serpent 2 Monte Carlo code

    International Nuclear Information System (INIS)

    Leppaenen, J.; Isotalo, A.

    2012-01-01

    This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

  1. A microcomputer program for coupled cycle burnup calculations

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Downar, T.J.; Taylor, E.L.

    1986-01-01

    A program, designated BRACC (Burnup, Reactivity, And Cycle Coupling), has been developed for fuel management scoping calculations, and coded in the BASIC language in an interactive format for use with microcomputers. BRACC estimates batch and cycle burnups for sequential reloads for a variety of initial core conditions, and permits the user to specify either reload batch properties (enrichment, burnable poison reactivity) or the target cycle burnup. Most important fuel management tactics (out-in or low-leakage loading, coastdown, variation in number of assemblies charged) can be simulated

  2. Safety aspects related to burnup increase and mixed oxide fuel

    International Nuclear Information System (INIS)

    Thomas, W.

    1992-01-01

    The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de

  3. Two dimensional burn-up calculation of TRIGA core

    International Nuclear Information System (INIS)

    Persic, A.; Ravnik, M.; Slavic, S.

    1996-01-01

    TRIGLAV is a new computer program for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport program WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. In this paper the calculation model of diffusion constants and burn-up calculation are described and some results of calculations for TRIGA MARK II reactor are presented. (author)

  4. Measurement and interpretation of triton burnup in Jet deuterium plasmas

    International Nuclear Information System (INIS)

    Jarvis, O.N.; Kallne, J.; Sadler, G.; van Belle, P.; Gorini, G.; Conroy, S.; Verschuur, K.

    1989-01-01

    The confinement and slowing down of fast tritons in JET deuterium plasmas is investigated. The ratio of 14 MeV and 2.5 MeV neutron production rates is measured. This ratio is equal to the fraction of tritons which burnup. The 2.5 MeV neutron emission is obtained from a set of fission chambers for which the calibration uncertainty is about 10%. The absolute calibration of the activation technique is calculated. The comparison between experimental and theoretical burnup ratios, for JET 1987 data, is shown. The range of conditions over which measurements of triton burnup fraction were obtained, is illustrated

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

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, M; Kitajima, S [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    1997-08-01

    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.

  6. IFPE/HBEP REV.1, Battelle's High Burn-Up Effects Programme for Fuel Performance

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2002-01-01

    Description: It contains data from phase 2 and 3 on fabrication, dimensions, fuel and cladding properties and composition, reactor conditions and Post Irradiation Examination (PIE) data of the High Burn-up Effects Programme (HBEP) carried out at the Battelle North-west Laboratories. Each data set contains a full irradiation history with clad temperature and local power listed for each rod at 5, 10 or 12 axial zones as a function of cumulative time to the end of the given time interval over which the power has been constant. Data is provided for 45 rods from phase 2 and 36 rods from phase 3. The different rods have been manufactured by: ASEA/TVO, BN, BNFL, FBFC, FRA/CEA, GE, KWU/CE, WEC

  7. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  8. The use of burnup credit for spent fuel cask design

    International Nuclear Information System (INIS)

    Lake, W.H.

    1993-01-01

    A new generation of high capacity spent fuel transport casks is being developed by the U.S. Department of Energy (DOE) as part of the Federal Waste Management System (FWMS). Burnup credit, which recognizes the reduced reactivity of spent fuel is being used for these casks. Two cask designs being developed for DOE by Babcock and Wilcox and General Atomics use burnup credit. The cask designs must be certified by the U.S. Nuclear Regulatory Commission (NRC) if they are to be used in the FWMS. Certification of these casks by the NRC would not require any change in the NRC's transport regulations, and would be consistent with past practices. Furthermore, use of burnup credit casks appears to be consistent with current International Atomic Energy Agency (IAEA) rules and regulations. To support NRC certification, DOE has identified the technical issues related to burnup credit, and embarked on a development program to resolve them. (J.P.N.)

  9. Burn-up measurement in the HTR-module-reactor

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

    The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

  10. Triton burnup measurements in KSTAR using a neutron activation system

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jungmin; Shi, Yue-Jiang; Chung, Kyoung-Jae, E-mail: jkjlsh1@snu.ac.k; Hwang, Y. S. [Department of Nuclear Engineering, Seoul National University, Seoul 151-744 (Korea, Republic of); Cheon, MunSeong; Rhee, T.; Kim, Junghee [National Fusion Research Institute, Daejeon 34133 (Korea, Republic of); Kim, Jun Young [Korea University of Science and Technology, Daejeon 34133 (Korea, Republic of); Isobe, M.; Ogawa, K. [National Institute for Fusion Science, Toki-shi (Japan); SOKENDAI (The Graduate University for Advanced Studies), Toki-shi (Japan)

    2016-11-15

    Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a {sup 3}He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%–0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

  11. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    Parks, C.V.

    1989-01-01

    The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of k eff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

  12. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

    2001-01-01

    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)

  13. Influence of FIMA burnup on actinides concentrations in PWR reactors

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

  14. TRIGA criticality experiment for testing burn-up calculations

    International Nuclear Information System (INIS)

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    1999-01-01

    A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)

  15. An optimal model for fuel burnup in nuclear reactors

    International Nuclear Information System (INIS)

    Anton, V.

    1979-05-01

    An approach to minimize the number of the burnup equations taking into account the introduction of an appropriate number of fission products is given. The corresponding number of fission pseudo-products is defined. (author)

  16. Impact of extended burnup on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-04-01

    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

  17. KENOREST - A new coupled code system based on KENO and OREST for criticality and burnup inventory calculations

    International Nuclear Information System (INIS)

    Hesse, U.; Gmal, B.; Voggenberger, Th.; Baleanu, M.; Langenbuch, S.

    2001-01-01

    The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)

  18. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

    2010-02-01

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  19. A guide introducing burnup credit, preliminary version. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)

  20. A burn-up module coupling to an AMPX system

    International Nuclear Information System (INIS)

    Salvatore Duque, M.; Gomez, S.E.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

    1990-01-01

    The Reactors and Neutrons Division of the Bariloche Atomic Center uses the AMPX system for the study of high conversion reactors (HCR). Such system allows to make neutronic calculations from the nuclear data library (ENDF/B-IV). The Nuclear Engineering career of the Balseiro Institute developed and implemented a burn-up module at a μ-cell level (BUM: Burn-up Module) which agrees with the requirement to be coupled to the AMPX system. (Author) [es

  1. Monte Carlo burnup simulation of the TAKAHAMA-3 benchmark experiment

    International Nuclear Information System (INIS)

    Dalle, Hugo M.

    2009-01-01

    High burnup PWR fuel is currently being studied at CDTN/CNEN-MG. Monte Carlo burnup code system MONTEBURNS is used to characterize the neutronic behavior of the fuel. In order to validate the code system and calculation methodology to be used in this study the Japanese Takahama-3 Benchmark was chosen, as it is the single burnup benchmark experimental data set freely available that partially reproduces the conditions of the fuel under evaluation. The burnup of the three PWR fuel rods of the Takahama-3 burnup benchmark was calculated by MONTEBURNS using the simplest infinite fuel pin cell model and also a more complex representation of an infinite heterogeneous fuel pin cells lattice. Calculations results for the mass of most isotopes of Uranium, Neptunium, Plutonium, Americium, Curium and some fission products, commonly used as burnup monitors, were compared with the Post Irradiation Examinations (PIE) values for all the three fuel rods. Results have shown some sensitivity to the MCNP neutron cross-section data libraries, particularly affected by the temperature in which the evaluated nuclear data files were processed. (author)

  2. Optimum Discharge Burnup and Cycle Length for PWRs

    International Nuclear Information System (INIS)

    Secker, Jeffrey R.; Johansen, Baard J.; Stucker, David L.; Ozer, Odelli; Ivanov, Kostadin; Yilmaz, Serkan; Young, E.H.

    2005-01-01

    This paper discusses the results of a pressurized water reactor fuel management study determining the optimum discharge burnup and cycle length. A comprehensive study was performed considering 12-, 18-, and 24-month fuel cycles over a wide range of discharge burnups. A neutronic study was performed followed by an economic evaluation. The first phase of the study limited the fuel enrichments used in the study to 235 U consistent with constraints today. The second phase extended the range of discharge burnups for 18-month cycles by using fuel enriched in excess of 5 wt%. The neutronic study used state-of-the-art reactor physics methods to accurately determine enrichment requirements. Energy requirements were consistent with today's high capacity factors (>98%) and short (15-day) refueling outages. The economic evaluation method considers various component costs including uranium, conversion, enrichment, fabrication and spent-fuel storage costs as well as the effect of discounting of the revenue stream. The resulting fuel cycle costs as a function of cycle length and discharge burnup are presented and discussed. Fuel costs decline with increasing discharge burnup for all cycle lengths up to the maximum discharge burnup considered. The choice of optimum cycle length depends on assumptions for outage costs

  3. Thermal property change of MOX and UO{sub 2} irradiated up to high burnup of 74 GWd/t

    Energy Technology Data Exchange (ETDEWEB)

    Nakae, Nobuo, E-mail: nakae-nobuo@jnes.go.jp [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Kurematsu, Shigeru; Kosaka, Yuji [Nuclear Development Corporation (NDC), 622-12, Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Yoshino, Aya; Kitagawa, Takaaki [Mitsubishi Nuclear Fuel Co., LTD. (MNF), 12-1, Yurakucho 1-Chome, Chiyoda-ku, Tokyo 100-0006 (Japan)

    2013-09-15

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO{sub 2} fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO{sub 2}. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO{sub 2} is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO{sub 2} at high burnup under the condition that the pellet–cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO{sub 2} before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO{sub 2}. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  4. Fission gas release at high burn-up: beyond the standard diffusion model

    International Nuclear Information System (INIS)

    Landskron, H.; Sontheimer, F.; Billaux, M.R.

    2002-01-01

    At high burn-up standard diffusion models describing the release of fission gases from nuclear fuel must be extended to describe the experimental loss of xenon observed in the fuel matrix of the rim zone. Marked improvements of the prediction of integral fission gas release of fuel rods as well as of radial fission gas profiles in fuel pellets are achieved by using a saturation concept to describe fission gas behaviour not only in the pellet rim but also as an additional fission gas path in the whole pellet. (author)

  5. Burnup degree measuring device for spent fuel

    International Nuclear Information System (INIS)

    Doi, Hideo; Imaizumi, Hideki; Endo, Yasumi; Itahara, Kuniyuki.

    1994-01-01

    The present invention provides a small-sized and convenient device for measuring a burnup degree of spent fuels, which can be installed without remodelling an existent fuel storage pool. Namely, a gamma-ray detecting portion incorporates a Cd-Te detector for measuring intensity ratio of gamma-rays. A neutron detecting portion incorporates a fission counter tube. The Cd-Te detector comprises a neutron shielding member for reducing radiation damages and a background controlling plate for reducing low energy gamma-rays entering from a collimator. Since the Cd-Td detector for use in a gamma-ray spectroscopy can be used at a normal temperature and can measure even a relatively strong radiation field, it can measure the intensity of gamma-rays from Cs-137 and Cs-134 in spent fuels accurately at a resolving power of less than 10 keV. Further, in a case where a cooling period is less than one year, gamma-rays from Rh-106 and Nb-95 can also be measured. (I.S.)

  6. Study on the sensitivity of Self-Powered Neutron Detectors (SPND) and its change due to burn-up

    International Nuclear Information System (INIS)

    Cho, Gyuseong; Lee, Wanno; Yoon, Jeong-Hyoun.

    1996-01-01

    Self-Powered Neutron Detectors (SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with its usage, it has some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation was accomplished to calculate the escape probability as a function of the birth position for the typical geometry of rhodium-based SPNDs. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity is calculated as a function of burn-up time in the reactor. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long-term usage. (author)

  7. Profiles

    International Nuclear Information System (INIS)

    2004-01-01

    Profiles is a synthetic overview of more than 100 national energy markets in the world, providing insightful facts and key energy statistics. A Profile is structured around 6 main items and completed by key statistics: Ministries, public agencies, energy policy are concerned; main companies in the oil, gas, electricity and coal sectors, status, shareholders; reserve, production, imports and exports, electricity and refining capacities; deregulation of prices, subsidies, taxes; consumption trends by sector, energy market shares; main energy projects, production and consumption prospects. Statistical Profiles are present in about 3 pages the main data and indicators on oil, gas, coal and electricity. (A.L.B.)

  8. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  9. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    International Nuclear Information System (INIS)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok

    2015-01-01

    The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235 U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233 U, 242 Pu, 150 Nd, and 133 Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code

  10. CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User's manual

    International Nuclear Information System (INIS)

    Serov, I.V.; Hoogenboom, J.E.

    1993-07-01

    The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or 'BuCo' material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP)

  11. Systemization of burnup sensitivity analysis code (2) (Contract research)

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

  12. Comparative study for axial and radial shuffling scheme effect on the performance of Pb-Bi cooled fast reactors with natural uranium as fuel cycle input

    International Nuclear Information System (INIS)

    Zaki Suud; Indah Rosidah; Maryam Afifah; Ferhat Aziz; Sekimoto, H.

    2013-01-01

    Full text:Comparative study for the Design of Pb-Bi cooled fast reactors with natural uranium as fuel cycle input using special radial shuffling strategy and axial direction modified CANDLE burn-up scheme has been performed. The reactors utilizes UN-PuN as fuel, Eutectic Pb-Bi as coolant, and can be operated without refueling for 10 years in each batch. Reactor design optimization is performed to utilize natural uranium as fuel cycle input. This reactor subdivided into 6-10 regions with equal volume in radial directions. The natural uranium is initially put in region 1, and after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions. The calculation has been done by using SRAC-Citation system code and JENDL-3.2 library. The effective multiplication factor change increases monotonously during 10 years reactor operation time. There is significant power distribution change in the central part of the core during the BOC and the EOC in the radial shuffling system. It is larger than that in the case of modified CANDLE case which use axial direction burning region move. The burn-up level of fuel is slowly grows during the first 15 years but then grow faster in the rest of burn-up history. This pattern is a little bit different from the case of modified CANDLE burn-up scheme in Axial direction in which the slow growing burn-up period is relatively longer almost half of the burn-up history. (author)

  13. High Frequency Acoustic Microscopy for the Determination of Porosity and Young's Modulus in High Burnup Uranium Dioxide Nuclear Fuel

    Science.gov (United States)

    Marchetti, Mara; Laux, Didier; Cappia, Fabiola; Laurie, M.; Van Uffelen, P.; Rondinella, V. V.; Wiss, T.; Despaux, G.

    2016-06-01

    During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. VR was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation.

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

    International Nuclear Information System (INIS)

    Clement Ravi Chandar, S.; Sivayya, D.N.; Puthiyavinayagam, P.; Chellapandi, P.

    2013-01-01

    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 B 4 C 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

  15. Use of burnup credit for transportation and storage

    International Nuclear Information System (INIS)

    Sanders, T.L.; Ewing, R.I.; Lake, W.H.

    1991-01-01

    Burnup credit is the application of the effects of fuel burnup to nuclear criticality design. When burnup credit is considered in the design of storage facilities and transportation casks for spent fuel, the objectives are to reduce the requirements for storage space and to increase the payload of casks with acceptable nuclear criticality safety margins. The spent-fuel carrying capacities of previous-generation transport casks have been limited primarily by requirements to remove heat and/or to provide shielding. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus a considerable weight margin is available to the designer for increasing the payload capacity. One method to achieve an increase in capacity is to reduce fuel assembly spacing. The amount of reduction in assembly spacing is limited by criticality and fuel support structural concerns. The optimum fuel assembly spacing provides the maximum cask loading within a basket that has adequate criticality control and sufficient structural integrity for regulatory accident scenarios. The incorporation of burnup credit in cask designs could result in considerable benefits in the transport of spent fuel. The acceptance of burnup credit for the design of transport casks depends on the resolution of system safety issues and the uncertainties that affect the determination of criticality safety margins. The remainder of this report will examine these issues and the integrated approach under way to resolve them. 20 refs., 2 figs

  16. Status of burnup credit implementation and research in Switzerland

    International Nuclear Information System (INIS)

    Grimm, P.

    2001-01-01

    Burnup credit has recently been approved by the Swiss licensing authority for the spent-fuel storage pool of a PWR plant for fuel exceeding the originally licensed initial enrichment. The criticality safety assessment is based on a configuration consisting of a small number (approximately a reload batch) of fresh assemblies surrounded by assemblies having a burnup corresponding to the minimum value in the top 1 m section after one cycle of irradiation. The allowable initial enrichment in this configuration is about 0.5% higher than for all fresh fuel. A central storage facility for all types of radioactive wastes from Switzerland, including cask storage of spent fuel assemblies is being commissioned presently. The first applications for licenses for casks to be used in this facility have been submitted. Credit for burnup has not been requested in these applications (conforming to the original licenses of the casks in their countries of origin), but utilities are interested in burnup credit for fuel with higher initial enrichments. Reactivity worth measurements as well as chemical assays of spent fuel samples in the LWR-PROTEUS facility at PSI are in detailed planning currently. The experiments, scheduled to start in 2001, will be performed in cooperation with the Swiss utilities and their fuel vendors. Although the focus of interest of these partners is on validation of in-core fuel management tools, the same experiments are also applicable to burnup credit, and contacts with further potential partners interested in this field are underway. (author)

  17. Advances in Metallic Fuels for High Burnup and Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

    2016-10-01

    Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

  18. Disposal criticality analysis methodology's principal isotope burnup credit

    International Nuclear Information System (INIS)

    Doering, T.W.; Thomas, D.A.

    2001-01-01

    This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)

  19. Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

    1996-12-01

    The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

  20. Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

    Energy Technology Data Exchange (ETDEWEB)

    Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki

    1996-12-01

    The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

  1. Liquid-metal fast breeder reactor fuel rod performance and modeling at high burnup

    International Nuclear Information System (INIS)

    Verbeek, P.; Toebbe, H.; Hoppe, N.; Steinmetz, B.

    1978-01-01

    The fuel rod modeling codes IAMBUS and COMETHE were used in the analysis and interpretation of postirradiation examination results of mixed-oxide fuel pins. These codes were developed in the framework of the SNR-300 research and development (R and D) program at Interatom and Belgonucleaire, respectively. SNR-300 is a liquid-metal fast breeder reactor demonstration plant designed and presently constructed in consortial cooperation by Germany, Belgium, and the Netherlands. RAPSODIE I, the two-bundle irradiation experiment, was irradiated in the French test FBR RAPSODIE FORTISSIMO and is one of the key irradiation experiments within the SNR-300 R and D program. The comparison of code predictions with postirradiation examination results concentrates on clad diameter expansions, clad total axial elongations, fuel differential and total axial elongations, fuel restructuring, and fission gas release. Fuel rod modeling was considered in the light of benchmarking of the codes, and there was consideration of fuel rod design for operation at low and high burnup

  2. On renormalization of axial anomaly

    International Nuclear Information System (INIS)

    Efremov, A.V.; Teryaev, O.V.

    1989-01-01

    It is shown that multiplicative renormalization of the axial singlet current results in renormalization of the axial anomaly in all orders of perturbation theory. It is a necessary condition for the Adler - Bardeen theorem being valid. 10 refs.; 2 figs

  3. Axial tomographic scanner

    International Nuclear Information System (INIS)

    1976-01-01

    An axial tomographic system is described comprising axial tomographic means for collecting sets of data corresponding to the transmission or absorption of a number of beams of penetrating radiation through a planar slice of an object. It includes means to locate an object to be analyzed, a source and detector for directing one or more beams of penetrating radiation through the object from the source to the detector, and means to rotate (and optionally translate) the source as well as means to process the collected sets of data. Data collection, data processing, and data display can each be conducted independently of each other. An additional advantage of the system described is that the raw data (i.e., the originally collected data) are not destroyed by the data processing but instead are retained intact for further reference or use, if needed

  4. New Sodium Cooled Long-Life Cores with Axially Multi-Driver Regions

    International Nuclear Information System (INIS)

    Hyun, Hae Ri; Hong, Ser Gi

    2014-01-01

    In this concept of long-life core (they are sometimes called B-B (Breed and Burn)), tall blanket is placed above the relatively short driver fuel. In the initial stage of burning, the power by fission is mostly generated in the driver region and it moves into the blanket region. The power and flux distributions that are highly peaked in the axial direction propagates slowly from the driver into the blanket region. This concept of long-life core fully utilizes the breeding of blanket in the fast spectra and it can achieve very high burnup of fuel. In this work, we introduce new sodium cooled longlife cores rating 600MWe (1800MWt). In these cores, the driver regions are heterogeneously placed into blanket region so as to achieve stabilized and less peaked axial power distribution as depletion proceeds. At present, our study is focused on only two axial driver regions but this concept can be easily extended onto the multi-driver region concept. The cores designed in this paper have two axial driver regions so as to have stabilized and less peaked axial power distributions as depletion proceeds. The results of the core design and analyses show that the cores have very long-lives longer than -49EFPYs and high discharge burnup higher than 200GWD/kg. Additionally, we considered a long-life core having no blanket. As expected, it was shown that these cores have stabilized and less peaked axial power distribution as the fuel depletes. However, the study shows that the cores having two driver regions still show high initial peaking of the axial power distributions and the core can be optimized by changing the driver fuel height

  5. TRIGA fuel element burnup determination by measurement and calculation

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.

    2000-01-01

    To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)

  6. Modelling of some high burnup phenomena in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, K; Lindstroem, F; Massih, A R [ABB Atom AB, Vaesteraas (Sweden)

    1997-08-01

    In this paper the results of some modelling efforts carried out by ABB Atom to describe certain light water reactor fuel high burnup effects are presented. In particular the degradation of fuel thermal conductivity with burnup and its impact on fuel temperature is briefly discussed. The formation of a porous rim and its effect on a thermal fission gas release has been modelled and the model has been used to predict the release of pressurized water reactor fuel rods that were operated at low power densities. Furthermore, a mathematical model which combines the diffusion and re-solution controlled thermal release with grain boundary movement has been briefly described. The model is used to compare release with diffusion only and release caused by diffusion and grain boundary sweeping (due to grain growth). Finally, analytical expressions are obtained for the calculation of fuel stoichiometry as a function of burnup. (author). 20 refs, 10 figs, 1 tab.

  7. Fuel cycle cost considerations of increased discharge burnups

    International Nuclear Information System (INIS)

    Scherpereel, L.R.; Frank, F.J.

    1982-01-01

    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

  8. Validation issues for depletion and criticality analysis in burnup credit

    International Nuclear Information System (INIS)

    Parks, C.V.; Broadhead, B.L.; Dehart, M.D.; Gauld, I.C.

    2001-01-01

    This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage, and disposal. The issues discussed are ones that have been identified by one or more constituents of the United States technical community (national laboratories, licensees, and regulators) that have been exploring the use of burnup credit. There is not necessarily agreement on the importance of the various issues, which sometimes is what creates the issue. The broad issues relate to the paucity of available experimental data (radiochemical assays and critical experiments) covering the full range and characteristics of spent nuclear fuel in away-from-reactor systems. The paper will also introduce recent efforts initiated at Oak Ridge National Laboratory (ORNL) to provide technical information that can help better assess the value of different experiments. The focus of the paper is on experience with validation issues related to use of burnup credit for transport and dry storage applications. (author)

  9. Preparation of higher-actinide burnup and cross section samples

    International Nuclear Information System (INIS)

    Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.

    1981-01-01

    A joint research program involving the United States and the United Kingdom was instigated about four years ago for the purpose of studying burnup of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of cross sections of a wide variety of higher actinide isotopes was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the burnup and cross section samples. The higher actinide samples chosen for the burnup study were 241 Am and 244 Cm in the forms of Am 2 O 3 , Cm 2 O 3 , and Am 6 Cm(RE) 7 O 21 , where (RE) represents a mixture of lanthanide sesquioxides. It is the purpose of this paper to describe technology development and its application in the preparation of the fuel specimens and the cross section specimens that are being used in this cooperative program

  10. Fission gas release from fuels at high burnup

    International Nuclear Information System (INIS)

    Kauffmann, Yves; Pointud, M.L.; Vignesoult, Nicole; Atabek, Rosemarie; Baron, Daniel.

    1982-04-01

    Determinations of residual gas concentrations by heating and by X microanalysis were respectively carried out on particles (TANGO program) and on sections of fuel rods, perfectly characterized as to fabrication and irradiation history. A threshold release temperature of 1250 0 C+-100 0 C was determined irrespective of the type of oxide and the irradiation history in the 18,000-45,000 MWdt -1 (U) specific burnup field. The overall analyses of gas released from the fuel rods show that, in the PWR operating conditions, the fraction released remains less than 1% up to a mean specific burnup of 35000 MWdt -1 (U). The release of gases should not be a limiting factor in the increase of specific burnups [fr

  11. Increased fuel burn-up and fuel cycle equilibrium

    International Nuclear Information System (INIS)

    Debes, M.

    2001-01-01

    Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

  12. COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System

    International Nuclear Information System (INIS)

    Suyama, K.; Masukawa, F.; Ido, M.; Enomoto, M.; Takyu, S.; Hara, T.

    2002-01-01

    1 - Description of program or function: Burn-up calculation of nuclear fuel. 2 - Methods: Matrix exponential method, Bateman Equation. 3 - Restrictions on the complexity of the problem: a) One-grouped cross section library should be prepared for the fuel system to be analyzed using UNITBURN. However, UNITBURN is not available now for UNIX systems. b) Gamma ray spectrometry calculation will fail using the attached piflib routine. This problem has already been rectified in the internal version. 4 - Typical running time: Two minutes for standard burn-up calculation on Sun ULTRA 30. 5 - Unusual features - a) Selection of Matrix exponential method, or Bateman Equation. b) JDDL, a detailed decay chain data based on ENSDF. 6 - Related or auxiliary programs: UNITBURN: Burnup calculation code unit cell system

  13. Role of measurement systems in burnup credit operations

    International Nuclear Information System (INIS)

    Ewing, R.I.; Sanders, T.L.

    1991-01-01

    Spent fuel transport casks designed using burnup credit have increased payloads that may greatly reduce the number of shipments required to transport spent fuel from reactor sites to repositories. Burnup credit is obtained by applying the reduced reactivity of spent fuel to considerations of nuclear criticality in the design of transport casks. Although it does not appear to be possible to directly measure the criticality of spent fuel assemblies, measurements can be employed to ensure that the only assemblies loaded into a cask have the characteristics appropriate to that cask design. An effective on-site measurement system must be matched to the characteristics of the spent fuel cask design and to the inventory of spent fuel. For operation reasons the system should be simple, accurate, efficient, and easily calibrated. This paper is part of a study to examine the effects of the spent fuel inventory in the U.S. on the selection of measurement systems useful in burnup credit operations

  14. Burnup credit implementation plan and preparation work at JAERI

    International Nuclear Information System (INIS)

    Nomura, Y.; Itahara, K.

    2001-01-01

    Application of the burnup credit concept is considered to be very effective to the design of spent fuel transport and storage facilities. This technology is all the more important when considering construction of the intermediate spent fuel storage facility, which is to be commissioned by 2010 due to increasing amount of accumulated spent fuel in Japan. Until reprocessing and recycling all the spent fuel arising, they will be stored as an energy stockpile until such time as they can be reprocessed. On the other hand, the burnup credit has been partly taken into account for the spent fuel management at Rokkasho Reprocessing Plant, which is to be commissioned in 2005. They have just finished the calibration tests for their burnup monitor with initially accepted several spent fuel assemblies. Because this monitoring system is employed with highly conservative safety margin, it is considered necessary to develop the more rational and simplified method to confirm burnup of spent fuel. A research program has been instituted to improve the present method employed at the spent fuel management system for the Spent Fuel Receiving and Storage Pool of Rokkasho Reprocessing Plant. This program is jointly performed by Japan Nuclear Fuel Limited (JNFL) and JAERI.This presentation describes the current status of spent fuel accumulation discharged from PWR and BWR in Japan and the recent incentive to introduce burnup credit into design of spent fuel storage and transport facilities. This also includes the content of the joint research program initiated by JNFL and JAERI. The relevant study has been continued at JAERI. The results by these research programs will be included in the Burnup Credit Guide Original Version compiled by JAERI. (author)

  15. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

    Melin, P.; Lavoine, O.; Houdaille, B.

    1986-04-01

    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 [fr

  16. Fission gas release from fuel at high burnup

    International Nuclear Information System (INIS)

    Meyer, R.O.; Beyer, C.E.; Voglewede, J.C.

    1978-03-01

    The release of fission gas from fuel pellets at high burnup is reviewed in the context of the safety analysis performed for reactor license applications. Licensing actions are described that were taken to correct deficient gas release models used in these safety analyses. A correction function, which was developed by the Nuclear Regulatory Commission staff and its consultants, is presented. Related information, which includes some previously unpublished data, is also summarized. The report thus provides guidance for the analysis of high burnup gas release in licensing situations

  17. Simulation of integral local tests with high-burnup fuel

    International Nuclear Information System (INIS)

    Gyori, G.

    2011-01-01

    The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

  18. Burnup measurements with the Los Alamos fork detector

    International Nuclear Information System (INIS)

    Bosler, G.E.; Rinard, P.M.

    1991-01-01

    The fork detector system can determine the burnup of spent-fuel assemblies. It is a transportable instrument that can be mounted permanently in a spent-fuel pond near a loading area for shipping casks, or be attached to the storage pond bridge for measurements on partially raised spent-fuel assemblies. The accuracy of the predicted burnup has been demonstrated to be as good as 2% from measurements on assemblies in the United States and other countries. Instruments have also been developed at other facilities throughout the world using the same or different techniques, but with similar accuracies. 14 refs., 2 figs., 2 tabs

  19. Isotopic biases for actinide-only burnup credit

    International Nuclear Information System (INIS)

    Rahimi, M.; Lancaster, D.; Hoeffer, B.; Nichols, M.

    1997-01-01

    The primary purpose of this paper is to present the new methodology for establishing bias and uncertainty associated with isotopic prediction in spent fuel assemblies for burnup credit analysis. The analysis applies to the design of criticality control systems for spent fuel casks. A total of 54 spent fuel samples were modeled and analyzed using the Shielding Analyses Sequence (SAS2H). Multiple regression analysis and a trending test were performed to develop isotopic correction factors for 10 actinide burnup credit isotopes. 5 refs., 1 tab

  20. Tag gas burnup based on three-dimensional FTR analysis

    International Nuclear Information System (INIS)

    Kidman, R.B.

    1976-01-01

    Flux spectra from a three-dimensional diffusion theory analysis of the Fast Test Reactor (FTR) are used to predict gas tag ratio changes, as a function of exposure, for each FTR fuel and absorber subassembly plenum. These flux spectra are also used to predict Xe-125 equilibrium activities in absorber plena in order to assess the feasibility of using Xe-125 gamma rays to detect and distinguish control rod failures from fuel rod failures. Worst case tag burnup changes are used in conjunction with burnup and mass spectrometer uncertainties to establish the minimum spacing of tags which allows the tags to be unambiguously identified

  1. Nuclear fuel burn-up economy; Ekonomija izgaranja nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-07-01

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

  2. Findings of an international study on burnup credit

    International Nuclear Information System (INIS)

    Brady, M.C.; Takano, M.; Okuno, H.; DeHart, M.D.; Nouri, A.

    1996-01-01

    Findings from a four year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses are presented in this paper. Approximately 20 participants from 11 countries have provided results for most problems. Four detailed benchmark problems for Pressurized Water Reactor (PWR) fuel have been completed and are summarized in this paper. Preliminary results from current work addressing burnup credit for Boiling Water Reactor (BWR) fuel will also be discussed as well as planned activities for additional benchmarks including Mixed-Oxide (MOX) fuels, subcritical benchmarks, international databases, and other activities

  3. Isotopic and criticality validation for actinide-only burnup credit

    International Nuclear Information System (INIS)

    Fuentes, E.; Lancaster, D.; Rahimi, M.

    1997-01-01

    The techniques used for actinide-only burnup credit isotopic validation and criticality validation are presented and discussed. Trending analyses have been incorporated into both methodologies, requiring biases and uncertainties to be treated as a function of the trending parameters. The isotopic validation is demonstrated using the SAS2H module of SCALE 4.2, with the 27BURNUPLIB cross section library; correction factors are presented for each of the actinides in the burnup credit methodology. For the criticality validation, the demonstration is performed with the CSAS module of SCALE 4.2 and the 27BURNUPLIB, resulting in a validated upper safety limit

  4. A regime showing anomalous triton burnup in JET

    International Nuclear Information System (INIS)

    Conroy, S.; Jarvis, O.N.; Sadler, G.; Pillon, M.

    1990-01-01

    Measurements of triton burnup made at JET in 1989 are in good agreement with a simple classical model of the triton slowing down, for the majority of discharges. For discharges with a long slowing down time (greater than 2 seconds), a much reduced burnup has been observed, suggesting that the tritons undergo diffusion with a diffusion constant of 0.10 m 2 s -1 . Also, the experimental 14 MeV neutron yield is 30% lower than expected for Beryllium limiter discharges. (author) 4 refs., 3 figs

  5. Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

    International Nuclear Information System (INIS)

    Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

    2001-01-01

    The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

  6. An axially and radially two-zoned large liquid-metal fast breeder reactor core concept

    International Nuclear Information System (INIS)

    Kamei, T.; Arie, K.; Moriki, Y.; Suzuki, M.; Yamaoka, M.

    1985-01-01

    A new core concept that has advantages over conventional homogeneous cores in neutronics characteristics such as power peaking factor, burnup reactivity loss, and reactivity response to the movement of control rods in earthquakes has been evolved. Two options of the new core concept are feasible. One is the so-called axially heterogeneous core, with the internal blanket placed at the lower part of the core. The other concept is similar to the conventional homogeneous core, but has two different plutonium-enriched zones in the axial as well as in the radial direction, so it is a hybrid type of the conventional homogeneous core and the axially heterogeneous core. The new design concept is described and the way that the core characteristics are improved by the chosen key parameters is shown

  7. Critical Axial Load

    Directory of Open Access Journals (Sweden)

    Walt Wells

    2008-01-01

    Full Text Available Our objective in this paper is to solve a second order differential equation for a long, simply supported column member subjected to a lateral axial load using Heun's numerical method. We will use the solution to find the critical load at which the column member will fail due to buckling. We will calculate this load using Euler's derived analytical approach for an exact solution, as well as Euler's Numerical Method. We will then compare the three calculated values to see how much they deviate from one another. During the critical load calculation, it will be necessary to calculate the moment of inertia for the column member.

  8. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  9. The Gd-isotopic fuel for high burnup in PWR's

    International Nuclear Information System (INIS)

    Dias, Marcio Soares; Mattos, João Roberto L. de; Andrade, Edison Pereira de

    2017-01-01

    Today, the discussion about the high burnup fuel is beyond the current fuel enrichment licensing and burnup limits. Licensing issues and material/design developments are again key features in further development of the LWR fuel design. Nevertheless, technological and economical solutions are already available or will be available in a short time. In order to prevent the growth of the technological gap, Brazil's nuclear sector needs to invest in the training of new human resources, in the access to international databases, and in the upgrading existing infrastructure. Experimental database and R&D infrastructure are essential components to support the autonomous development of Brazilian Nuclear Reactors, promoting the development of national technologies. The (U,Gd)O_2 isotopic fuel proposed by the CDTN's staff solve two main issues in the high burnup fuel, which are (1) the peak of reactivity resulting from the Gd-157 fast burnup, and (2) the peak of temperature in the (U,Gd)O_2 nuclear fuel resulting from detrimental effects in the thermal properties for gadolinia additions higher than 2%. A sustainable future can be envisaged for the nuclear energy. (author)

  10. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

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

    International Nuclear Information System (INIS)

    Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.

    2006-01-01

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

  12. Determination of reactor fuel burnup using passive neutron assay

    International Nuclear Information System (INIS)

    Kodeli, I.; Trkov, A.; Najzer, M.; Ertek, C.

    1988-01-01

    Passive neutron assay (PNA) method was developed to verify the fissile inventory of the irradiated reactor fuels. The characteristics of the method were studied at 'Jozef Stefan' Institute. The dependence of neutron source in the fuel on burnup, cooling time, initial enrichment and specific power were investigated and the accuracy of the method, using available computer codes was estimated. (author)

  13. Burnup measurement study and prototype development in HTR-PM

    International Nuclear Information System (INIS)

    Yan Weihua; Zhang Zhao; Xiao Zhigang; Zhang Liguo

    2014-01-01

    In a pebble-bed core which employs the multi-pass scheme, it is mandatory to determine the burnup of each pebble after the pebble has been extracted from the core in order to determine whether its design burnup has been reached or whether it has to be reinserted into the core again. The burnup of the fuel pebbles can be determined by measuring the activity of 137 Cs with an HPGe detector because of their good correspondence, which is independent of the irradiation history in the core. Based on experiments and Geant4 simulation, the correction factor between the fuel and calibration source was derived by using the efficiency transfer method. By optimizing spectrum analysis algorithm and parameters, the relative standard deviation of the 137 Cs activity can be still controlled below 3.0% despite of the presence of interfering peaks. On the foundation of the simulation and experiment research, a complete solution for burnup measurement system in HTR-PM is provided. (authors)

  14. Extended burnup with SEU fuel in Atucha-1 NPP

    International Nuclear Information System (INIS)

    Alvarez, L.; Casario, J.; Fink, J.; Perez, R.; Higa, M.

    2002-01-01

    Atucha-1 is a Pressurized Heavy Water Reactor originally fuelled with natural uranium. Fuel Assemblies consist of 36 fuel rods and the active length is 5300 mm. The total length of the fuel assembly is about 6 m. The average discharge burnup of natural UO 2 fuel is 5900 MWd/tU. After the deregulation of the Argentine electricity market there was an important incentive to reduce the impact of fuel cost on the cost of generation. To keep the competitiveness of the nuclear energy against another sources of electricity it was necessary to reduce the cost of the nuclear fuel. With this objective a program to introduce SEU (0.85 % 235 U) fuel in Atucha-1 was launched in 1993. As a result of this program the average SEU fuel discharge burnup increased to more than 11000 MWd/tU. The first SEU fuels were introduced in Atucha-1 in 1995 and, in the present stage of the program, 71% of core positions are loaded with this type of fuel. This paper describes key aspects of Atucha-1 fuel design and their relevance limiting the burnup extension and shows relevant data regarding the SEU in-reactor performance. At the present time 125 SEU Fuel Assemblies have been irradiated without failures associated with the extended burnup or unfavorable influences on the operation of the power station. (author)

  15. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  16. UABUC - Single energy point model burnup computer code for water reactors

    International Nuclear Information System (INIS)

    El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

    1981-01-01

    UABUC is a single energy point reactor burnup computer program in FORTRAN language. The program calculates the change in the isotopic composition of the uranium fuel as a function of irradiation time with all its associated quantities such as the average point flux, the conversion ratio, macroscopic fuel cross sections, and the point reactivity profile. A step-wise time analytical solution was developed for the nonlinear first order burnup differential equations. The ''Westcott'' convention of the effective cross sections was used except for plutonium-240 and uranium-238. For plutonium-240, an effective microscopic cross section was derived from the direct physical arguments taking into account the selfshielding effect of plutonium-240 as well as the 1 ev. resonance absorption. For uranium-238, an effective cross section, reflecting the effect of fast fission and resonance absorption was used. The fission products were treated in the three groups with 50, 300, and 800 barns. The yields in the groups were treated as functions of the type of fissionable nuclides, the effective neutron temperature, and the epithermal index. Xenon-135 and Samarium-149 were treated separately as functions of irradiation time. (author)

  17. M5TM alloy high burnup behavior and worldwide licensing

    International Nuclear Information System (INIS)

    Mardon, J.P.; Hoffmann, P.B.; Garner, G.L.

    2005-01-01

    The in-reactor behavior of advanced PWR Zirconium alloys at burnups equal to or below licensing limits has been widely reported. Specifically, the advanced alloy M5 has demonstrated impressive improvements over Zircaloy-4 for fuel rod cladding and fuel assembly structural components. To demonstrate superiority of the alloy at burnups beyond current licensing limits, M5 has been operated in PWR at burnups exceeding 71 GWd/tU in the United States and 78 GWd/tU in Europe. Two extensive irradiation programs have been performed in the United States to demonstrate alloy M5 performance beyond current licensing limits. Four M5 TM fuel rods were exposed to four 24-month cycles in a 15x15 reactor beginning in 1995. Additionally, one 17x17 lead assembly containing M5 fuel rods and guide tubes was operated for four 18-month cycles beginning from 1997. Post-irradiation examinations (PIE) performed after all four cycles in the 15x15 demonstration program revealed excellent performance in the licensed burnup and in the high burnup stages of the experience. Examination of the 4th cycle 17x17 assembly will be accomplished in two stages the first of which is scheduled for June 2005. Moreover, several irradiation campaigns have been performed in Europe in order to confirm the excellent M5 in-pile behavior in demanding PWRs irradiation conditions with regard to void fraction, heat flux, lithium content and temperature. Results from the high burnup fuel examinations verify that the excellent performance achieved up to 62 GWd/tU was continued into higher burnup. The results of high burnup PIE campaigns for European and American PWR's are presented in this paper. Measured performance indicators include fuel assembly dimensional stability parameters (assembly length, fuel rod length, assembly bow, fuel rod bow, fuel rod radial creep and spacer grid width), oxidation measurements (fuel rod and guide tube) and hydrogen pick-up data (fuel rod). In the framework of PCI studies, power ramp

  18. Comparison of analysis methods for burnup credit applications

    International Nuclear Information System (INIS)

    Sanders, T.L.; Brady, M.C.; Renier, J.P.; Parks, C.V.

    1989-01-01

    The current approach used for the development and certification of spent fuel storage and transport casks requires an assumption of fresh fuel isotopics in the criticality safety analysis. However, it has been shown that there is a considerable reactivity reduction when the isotopics representative of the depleted (or burned) fuel are used in a criticality analysis. Thus, by taking credit for the burned state of the fuel (i.e., burnup credit), a cask designer could achieve a significant increase in payload. Accurate prediction of k eff for spent fuel arrays depends both on the criticality safety analysis and the prediction of the spent fuel isotopics via a depletion analysis. Spent fuel isotopics can be obtained from detailed multidimensional reactor analyses, e.g. the code PDQ, or from point reactor burnup models. These reactor calculations will help verify the adequacy of the isotopics and determine Δk eff biases for various analysis assumptions (with and without fission products, actinide absorbers, burnable poison rods, etc.). New software developed to interface PDQ multidimensional isotopics with KENO V.a reactor and cask models is described. Analyses similar to those performed for the reactor cases are carried out with a representative burnup credit cask model using the North Anna fuel. This paper presents the analysis methodology that has been developed for evaluating the physics issues associated with burnup credit. It is applicable in the validation and characterization of fuel isotopics as well as in determining the influence of various analysis assumptions in terms of δk eff . The methodology is used in the calculation of reactor restart criticals and analysis of a typical burnup credit cask

  19. Reactivity effect of spent fuel due to spatial distributions for coolant temperature and burnup

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Yamane, Y. [Nagoya Univ., Dept. of Nuclear Engineering, Nagoya, Aichi (Japan); Suyama, K. [OECD/NEA, Paris (France); Mochizuki, H. [Japan Research Institute, Ltd., Tokyo (Japan)

    2002-03-01

    We investigated the reactivity effect of spent fuel caused by the spatial distributions of coolant temperature and burnup by using the integrated burnup calculation code system SWAT. The reactivity effect which arises from taking account of the spatial coolant temperature distribution increases as the average burnup increases, and reaches the maximum value of 0.69%{delta}k/k at 50 GWd/tU when the burnup distribution is concurrently considered. When the burnup distribution is ignored, the reactivity effect decreases by approximately one-third. (author)

  20. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  1. Axial skeletal CT densitometry

    International Nuclear Information System (INIS)

    Lampmann, L.E.H.

    1982-01-01

    Since the discovery of the Roentgen ray a precise and accurate assessment of bone mineral content has been a challenge to many investigators. A number of methods have been developed but no one satisfied. Considering its technical possibilities computed tomography is very promising in determination of bone mineral content (BMC). The new modality enables BMC estimations in the axial skeletal trabecular bone. CT densitometry can be performed on a normal commercially available third generation whole body CT scanner. No dedicated device in a special clinical set-up is necessary. In this study 106 patients, most of them clinically suspected of osteoporosis, were examined. The new method CT densitometry has been evaluated. The results have been correlated to alternative BMC determination methods. (Auth.)

  2. Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

    International Nuclear Information System (INIS)

    Tohjoh, Masayuki; Endo, Tomohiro; Watanabe, Masato; Yamamoto, Akio

    2006-01-01

    As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore, if statistical error alone is evaluated, the errors in Monte Carlo burn-up calculations may be underestimated. To make clear this effect of error propagation on Monte Carlo burn-up calculations, we here proposed an equation that can predict the variance of nuclide number densities after burn-up calculations, and we verified this equation using enormous numbers of the Monte Carlo burn-up calculations by changing only the initial random numbers. We also verified the effect of the number of burn-up calculation points on Monte Carlo burn-up calculations. From these verifications, we estimated the errors in Monte Carlo burn-up calculations including both statistical and propagated errors. Finally, we made clear the effects of error propagation on Monte Carlo burn-up calculations by comparing statistical errors alone versus both statistical and propagated errors. The results revealed that the effects of error propagation on the Monte Carlo burn-up calculations of 8 x 8 BWR fuel assembly are low up to 60 GWd/t

  3. An economic evaluation of a storage system for casks with burnup credit

    International Nuclear Information System (INIS)

    Mimura, Masahiro; Tsuda, Kazuaki; Yamada, Nobuyuki; O-iwa, Akio.

    1993-01-01

    It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

  4. Configuration of LWR fuel enrichment or burnup yielding maximum power

    International Nuclear Information System (INIS)

    Bartosek, V.; Zalesky, K.

    1976-01-01

    An analysis is given of the spatial distribution of fuel burnup and enrichment in a light-water lattice of given dimensions with slightly enriched uranium, at which the maximum output is achieved. It is based on the spatial solution of neutron flux using a one-group diffusion model in which linear dependence may be expected of the fission cross section and the material buckling parameter on the fuel burnup and enrichment. Two problem constraints are considered, i.e., the neutron flux value and the specific output value. For the former the optimum core configuration remains qualitatively unchanged for any reflector thickness, for the latter the cases of a reactor with and without reflector must be distinguished. (Z.M.)

  5. MCB. A continuous energy Monte Carlo burnup simulation code

    International Nuclear Information System (INIS)

    Cetnar, J.; Wallenius, J.; Gudowski, W.

    1999-01-01

    A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (author)

  6. OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN

    International Nuclear Information System (INIS)

    Hesse, Ulrich; Sieberer, Johann

    2006-01-01

    1 - Description of program or function: In OREST, the 1-dimensional lattice code HAMMER and the isotope generation and depletion code ORIGEN are directly coupled for burnup simulation in light-water reactor fuels (GRS recommended). Additionally heavy water and graphite moderated systems can be calculated. New version differs from the previous version in the following features: An 84-group-library LIB84 for up to 200 isotopes is used to update the 3-group -POISON-XS. LIB84 uses the same energy boundaries as THERMOS and HAMLET in . In this way, high flexibility is achieved in very different reactor models. The coupling factor between THERMOS and HAMLET is now directly transferred from HAMMER to THERES and omits the equation 4 (see page 6 of the manual). Sandwich-reactor fuel reactivity and burnup calculations can be started with NGEOM = 1. Thorium graphite reactivity and burnup calculations can be started with NLIBE = 1. High enriched U-235 heavy water moderated reactivity and burnup calculations can be started. HAMLET libraries in for U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-242, Am-241, Am-243 and Zirconium are updated using resonance parameters. NEA-1324/04: A new version of the module hamme97.f has replaced the old one. 2 - Method of solution: For the user-defined irradiation history, an input data processor generates program loops over small burnup steps for the main codes HAMMER and ORIGEN. The user defined assembly description is transformed to an equivalent HAMMER fuel cell. HAMMER solves the integral neutron transport equation in a four-region cylindrical or sandwiched model with reflecting boundaries and runs with fuel power calculated rod temperatures. ORIGEN runs with HAMMER-calculated cross sections and neutron spectra and calculates isotope concentrations during burnup by solving the buildup-, depletion- and decay-chain equations. An output data processor samples the outputs of the program modules and generates tabular works for the

  7. Nuclide Importance and the Steady-State Burnup Equation

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nemoto, Atsushi

    2000-01-01

    Conventional methods for evaluating some characteristic values of nuclides relating to burnup in a given neutron spectrum are reviewed in a mathematically systematic way, and a new method based on the importance theory is proposed. In this method, these characteristic values of a nuclide are equivalent to the importances of the nuclide. By solving the equation adjoint to the steady-state burnup equation with a properly chosen source term, the importances for all nuclides are obtained simultaneously.The fission number importance, net neutron importance, fission neutron importance, and absorbed neutron importance are evaluated and discussed. The net neutron importance is a measure directly estimating neutron economy, and it can be evaluated simply by calculating the fission neutron importance minus the absorbed neutron importance, where only the absorbed neutron importance depends on the fission product. The fission neutron importance and absorbed neutron importance are analyzed separately, and detailed discussions of the fission product effects are given for the absorbed neutron importance

  8. CARA design criteria for HWR fuel burnup extension

    International Nuclear Information System (INIS)

    Florido, P.C.; Cirimello, R.O.; Bergallo, J.E.; Marino, A.C.; Delmastro, D.F.; Brasnarof, D.O.; Gonzalez, J.H.; Juanico, L.A.

    2002-01-01

    A new concept for HWR fuel bundles, namely CARA, is presented. The CARA design allows to improve all the major performances in the PHWR fuel technology. Among others, it reaches higher burnup and thermohydraulic safety margins, together with lower fuel pellet temperatures and Zry/HM mass ratio. Moreover, it keeps the fuel mass content per unit length and the channel pressure drop by using a single diameter of fuel rods. (author)

  9. Effect of fuel burnup on the mechanical safety coefficients

    International Nuclear Information System (INIS)

    Plyashkevich, V.Ju.; Sidorenko, V.D.; Shishkov, L.K.

    2001-01-01

    )In the paper the results of studies of changes in the process of campaign 'disturbances' of local heat flux and local fuel burnup, resulting from the 'mechanical' deviations in the composition and geometrical characteristics of fuel rods from the nominal are given. As example, the WWER-440 fuel assembly with burnable poisons used in the five-year fuel cycle is considered. The effect of deviations in fuel enrichment, fuel content, gadolinium content and geometrical size was studied (Authors)

  10. Validation of SCALE-4 for burnup credit applications

    International Nuclear Information System (INIS)

    Bowman, S.M.; DeHart, M.D.; Parks, C.V.

    1995-01-01

    In the past, a criticality analysis of PWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at ORNL in support of the US DOE efforts to demonstrate a validation approach for criticality safety methods to be used in burnup credit cask design. To date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The ANSI/ANS-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors by correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

  11. Burnup calculations for cadmium. A case study for HFR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Sciolla, C.M

    2000-09-11

    This report describes the pre-design burnup calculations performed for a cadmium shielded high fluence irradiation experiment in the HFR. The very high absorption cross section in cadmium causes problems in the calculations for two different reasons. Firstly, because of the large reaction rates the assumption that the flux and the cross sections remain piecewise constant is no longer true. Therefore the correct solution can only be obtained when using extremely small time steps which leads to excessive computing times. Secondly, the self-shielding in the cadmium becomes complete (black absorber) causing the depletion to progress in a shell-wise manner. As a consequence the depletion evolves nearly linear instead of exponential with time. Because of this the depletion codes are used in a regime for which these have not been designed leading to a systematic error. The analysis shows however that a good estimate for the burnup time can be obtained by extrapolation from calculations with practically sized time steps and a correction is derived to compensate the systematic error. The calculations were done using the OCTOPUS burnup code system, including the 3-D Monte-Carlo spectrum code MCNP-4B and the depletion code FISPACT-4.2. Verifications were performed with the WIMS code system. The first part of the report describes the study of the cadmium burnup calculations for a shielded steel sample with the emphasis on analyzing the requirements for obtaining the correct solution. The second part describes the time-dependent power production calculations with the steel replaced by lithium containing ceramic material such as to be used in the 'High Fluence Irradiation of Ceramics for Fusion' (HICU) experiment. 12 refs.

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

    International Nuclear Information System (INIS)

    White, J.R.

    1979-01-01

    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

  13. Calculation of triton confinement and burn-up in tokamaks

    International Nuclear Information System (INIS)

    Anderson, D.; Battistoni, P.

    1987-01-01

    An analytical investigation is made of the confinement and subsequent burn-up of fusion produced tritons in a deuterium Tokamak plasma. Explicit approximations are obtained for the triton confinement factor, clearly displaying the scaling with physical parameters. The importance of pitch angle scattering losses during the triton slowing down is also estimated. A comparison with experiments and numerical calculations on the FT Tokamak slows good qualitative agreement. (authors)

  14. Potential of axial fuel management strategies in thorium-fuelled CANDU's

    International Nuclear Information System (INIS)

    Milgram, M.S.

    1978-06-01

    Three axial fuel management strategies are compared for use in a CANDU-PHW reactor operating on a self-sufficient, equilibrium thorium cycle. Two of these strategies are familiar ones for uranium reactors, and the third seeks to take advantage of the nuclear characteristics of the Th 232 → U 233 transmutation chain to improve the economics of the fuel cycle by periodically removing the fuel from the reactor. This results in an approximately 50% increase in burnup and an approximately 15% decrease in heavy element fuel inventory at a channel power of 6 MW, relative to the other strategies. (author)

  15. BASHAN: A few-group three-dimensional diffusion code with burnup and fuel management features

    International Nuclear Information System (INIS)

    Pearce, D.F.

    1970-12-01

    The diffusion equation for a two or three-dimensional, two-group or multi-group downscatter problem is solved by conventional finite difference techniques. An x-y-z geometry is assumed with an 'in-channel' mesh point representation. Options are available which allow representation of a soluble poison dispersed throughout the reactor, and also absorber rods in specified channels. The power distribution and multiplication factor k eff are calculated and a point rating map is used to advance the irradiation at each mesh point by a specified time-step so that burnup is followed. Fuel changes may be made so that radial shuffling and axial shuffling fuel management schemes can be studies. The code has been written in FORTRAN S2 for an IBM 7030 (STRETCH) computer which, with a fast store of 80,000 locations, allows problems of up to 15,000 mesh points to be dealt with. Conversion to FORTRAN IV for IBM 360 has now been completed. (author)

  16. Dissipative Axial Inflation

    CERN Document Server

    Notari, Alessio

    2016-12-22

    We analyze in detail the background cosmological evolution of a scalar field coupled to a massless abelian gauge field through an axial term $\\frac{\\phi}{f_\\gamma} F \\tilde{F}$, such as in the case of an axion. Gauge fields in this case are known to experience tachyonic growth and therefore can backreact on the background as an effective dissipation into radiation energy density $\\rho_R$, which which can lead to inflation without the need of a flat potential. We analyze the system, for momenta $k$ smaller than the cutoff $f_\\gamma$, including numerically the backreaction. We consider the evolution from a given static initial condition and explicitly show that, if $f_\\gamma$ is smaller than the field excursion $\\phi_0$ by about a factor of at least ${\\cal O} (20)$, there is a friction effect which turns on before that the field can fall down and which can then lead to a very long stage of inflation with a generic potential. In addition we find superimposed oscillations, which would get imprinted on any kind of...

  17. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.; DeHart, M.D.

    1999-01-01

    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

  18. Application of burnup credit for PWR spent fuel storage pool

    International Nuclear Information System (INIS)

    Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

    1999-01-01

    A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

  19. Three dimensional Burn-up program parallelization using socket programming

    International Nuclear Information System (INIS)

    Haliyati R, Evi; Su'ud, Zaki

    2002-01-01

    A computer parallelization process was built with a purpose to decrease execution time of a physics program. In this case, a multi computer system was built to be used to analyze burn-up process of a nuclear reactor. This multi computer system was design need using a protocol communication among sockets, i.e. TCP/IP. This system consists of computer as a server and the rest as clients. The server has a main control to all its clients. The server also divides the reactor core geometrically to in parts in accordance with the number of clients, each computer including the server has a task to conduct burn-up analysis of 1/n part of the total reactor core measure. This burn-up analysis was conducted simultaneously and in a parallel way by all computers, so a faster program execution time was achieved close to 1/n times that of one computer. Then an analysis was carried out and states that in order to calculate the density of atoms in a reactor of 91 cm x 91 cm x 116 cm, the usage of a parallel system of 2 computers has the highest efficiency

  20. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    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

  1. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  2. CEA contribution to power plant operation with high burnup level

    International Nuclear Information System (INIS)

    1981-03-01

    High level burnup in PWR leads to investigate again the choices carried out in the field of fuel management. French CEA has studied the economic importance of reshuffling technique, cycle length, discharge burnup, and non-operation period between two cycles. Power plants operators wish to work with increased length cycles of 18 months instead of 12. That leads to control problems because the core reactivity cannot be controlled with the only soluble boron: moderator temperature coefficient must be negative. With such cycles, it is necessary to use burnable poisons and for economic reasons with a low penalty in end of cycle. CEA has studied the use of Gd 2 O 3 mixed with fuel or with inert element like Al 2 O 3 . Parametric studies of specific weights, efficacities relatively to the fuel burnup and the fuel enrichment have been carried out. Particular studies of 1 month cycles with Gd 2 O 3 have shown the possibility to control power distribution with a very low reactivity penalty in EOC. In the same time, in the 100 MW PWR-CAP, control reactivity has been made with large use of gadolinia in parallel with soluble boron for the two first cycles

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

    International Nuclear Information System (INIS)

    Proselkov, V.; Saprykin, V.; Scheglov, A.

    2003-01-01

    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

  4. Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5.5 at. % burnup

    International Nuclear Information System (INIS)

    Strain, R.V.; Johnson, C.E.

    1978-05-01

    Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760 0 C. The maximum diametral change that occurred during irradiation was 0.2% ΔD/D 0 . The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred

  5. Bessel beam CARS of axially structured samples

    Science.gov (United States)

    Heuke, Sandro; Zheng, Juanjuan; Akimov, Denis; Heintzmann, Rainer; Schmitt, Michael; Popp, Jürgen

    2015-06-01

    We report about a Bessel beam CARS approach for axial profiling of multi-layer structures. This study presents an experimental implementation for the generation of CARS by Bessel beam excitation using only passive optical elements. Furthermore, an analytical expression is provided describing the generated anti-Stokes field by a homogeneous sample. Based on the concept of coherent transfer functions, the underling resolving power of axially structured geometries is investigated. It is found that through the non-linearity of the CARS process in combination with the folded illumination geometry continuous phase-matching is achieved starting from homogeneous samples up to spatial sample frequencies at twice of the pumping electric field wave. The experimental and analytical findings are modeled by the implementation of the Debye Integral and scalar Green function approach. Finally, the goal of reconstructing an axially layered sample is demonstrated on the basis of the numerically simulated modulus and phase of the anti-Stokes far-field radiation pattern.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

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

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

    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. 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 ∞ 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)

  8. 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)

  9. CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I V; Hoogenboom, J E

    1993-07-01

    The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or `BuCo` material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP).

  10. Characterization of Multiflux Axial Compressors

    International Nuclear Information System (INIS)

    Brasnarof, Daniel; Kyung Kyu-Hyung; Rivarola, Martin; Gonzalez Jose; Florido, Pablo; Orellano, Pablo; Bergallo, Juan

    2003-01-01

    In the present work the results of analytical models of performance are compared with experimental data acquired in the multi flux axial compressor test facility, built in The Pilcaniyeu Technological Complex for the SIGMA project.We describe the experimental circuit and the data of the dispersion inside the axial compressor obtained using a tracer gas through one of the annular inlets.The attained results can be used to validate the design code for the multi flux axial compressors and SIGMA industrial plant

  11. Axial gap rotating electrical machine

    Science.gov (United States)

    None

    2016-02-23

    Direct drive rotating electrical machines with axial air gaps are disclosed. In these machines, a rotor ring and stator ring define an axial air gap between them. Sets of gap-maintaining rolling supports bear between the rotor ring and the stator ring at their peripheries to maintain the axial air gap. Also disclosed are wind turbines using these generators, and structures and methods for mounting direct drive rotating electrical generators to the hubs of wind turbines. In particular, the rotor ring of the generator may be carried directly by the hub of a wind turbine to rotate relative to a shaft without being mounted directly to the shaft.

  12. Study of axial magnetic effect

    Energy Technology Data Exchange (ETDEWEB)

    Braguta, Victor [IHEP, Protvino, Moscow region, 142284 Russia ITEP, B. Cheremushkinskaya street 25, Moscow, 117218 (Russian Federation); School of Biomedicine, Far Eastern Federal University, Ajax 10 Building 25, Russian island, Vladivostok, 690922 (Russian Federation); Chernodub, M. N. [CNRS, Laboratoire de Mathématiques et Physique Théorique, Université François-Rabelais Tours, Fédération Denis Poisson, Parc de Grandmont, 37200 Tours, France Department of Physics and Astronomy, University of Gent, Krijgslaan 281, S9, B-9000 Gent (Belgium); School of Biomedicine, Far Eastern Federal University, Ajax 10 Building 25, Russian island, Vladivostok, 690922 (Russian Federation); Goy, V. A. [School of Natural Sciences, Far Eastern Federal University, Sukhanova street 8, Vladivostok, 690950 (Russian Federation); Landsteiner, K. [Instituto de Física Teórica UAM/CSIC, C/ Nicolás Cabrera 13-15, Universidad Autónoma de Madrid, Cantoblanco, 28049 Madrid (Spain); Molochkov, A. V. [School of Biomedicine, Far Eastern Federal University, Ajax 10 Building 25, Russian island, Vladivostok, 690922 (Russian Federation); Ulybyshev, M. [ITEP, B. Cheremushkinskaya street 25, Moscow, 117218 Russia Institute for Theoretical Problems of Microphysics, Moscow State University, Moscow, 119899 (Russian Federation)

    2016-01-22

    The Axial Magnetic Effect manifests itself as an equilibrium energy flow of massless fermions induced by the axial (chiral) magnetic field. Here we study the Axial Magnetic Effect in the quenched SU(2) lattice gauge theory with massless overlap fermions at finite temperature. We numerically observe that in the low-temperature hadron phase the effect is absent due to the quark confinement. In the high-temperature deconfinement phase the energy flow is an increasing function of the temperature which reaches the predicted asymptotic T{sup 2} behavior at high temperatures. We find, however, that energy flow is about one order of magnitude lower compared to a theoretical prediction.

  13. Fundamental burn-up mode in a pebble-bed type reactor

    International Nuclear Information System (INIS)

    Chen, Xue-Nong; Kiefhaber, Edgar; Maschek, Werner

    2008-01-01

    This paper deals with a pebble-bed type reactor, in which the fuel is loaded from one side (top) and discharged from the other side (bottom). A boundary value problem of a single group diffusion equation coupled with simplified burn-up equations is studied, where the natural radioactive decay processes are neglected in the burn-up modelling. An asymptotic burning wave solution is found analytically in the one-dimensional case, which is called as fundamental burn-up mode. Among this solution family there are two particular cases, namely, a classic fundamental solution with a zero burn-up and a partial solitary burn-up wave solution with a highest burn-up. An example of Th-U conversion is considered and the solutions are presented in order to show the mechanism of the burning wave. (author)

  14. Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz

    2000-01-01

    This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

  15. Polynomial expansion methodology for microscopic cross sections to use in spatial burnup calculations

    International Nuclear Information System (INIS)

    Conti Filho, P.; Oliveira Barroso, A.C. de

    1985-01-01

    It was developed a computer code to generate polynomial coefficients which represent homogenized microscopic cross sections in function of the local accumulated burnup and concentration of soluble boron, presented in fuel element, for each step of burnup reactor. Afterward, it was developed a coupling between LEOPARD-GERADOR DE POLINOMIOS - CITATION computer codes to interpret and build homogenized microscopic cross sections according with local characteristics of each fuel element during the burnup calculation of reactor core. (M.C.K.) [pt

  16. Comparison of measured and calculated burn-up of AVR-Fuel-Elements

    Energy Technology Data Exchange (ETDEWEB)

    Wagemann, R.

    1974-03-15

    Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

  17. Effect of local burn-up variation on computed mean nuclide concentrations

    International Nuclear Information System (INIS)

    Moeller, W.

    1982-01-01

    Mean concentrations of U-235, U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 in some volume areas of WWER-440 fuel assemblies have been calculated from corresponding burn-up microdistribution data and compared with those calculated from burn-up mean values. Differences occurring were below 3% for the uranium nuclides but, at low burn-ups, considerable for Pu-241 and Pu-242. (author)

  18. Dissipative axial inflation

    Energy Technology Data Exchange (ETDEWEB)

    Notari, Alessio [Departament de Física Fondamental i Institut de Ciències del Cosmos, Universitat de Barcelona, Martí i Franquès 1, Barcelona, 08028 Spain (Spain); Tywoniuk, Konrad, E-mail: notari@ffn.ub.es, E-mail: konrad.tywoniuk@cern.ch [Theoretical Physics Department, CERN, Geneva (Switzerland)

    2016-12-01

    We analyze in detail the background cosmological evolution of a scalar field coupled to a massless abelian gauge field through an axial term φ/ f {sub γ} F ∼ F , such as in the case of an axion. Gauge fields in this case are known to experience tachyonic growth and therefore can backreact on the background as an effective dissipation into radiation energy density ρ{sub R}, which can lead to inflation without the need of a flat potential. We analyze the system, for momenta k smaller than the cutoff f {sub γ}, including the backreaction numerically. We consider the evolution from a given static initial condition and explicitly show that, if f {sub γ} is smaller than the field excursion φ{sub 0} by about a factor of at least O (20), there is a friction effect which turns on before the field can fall down and which can then lead to a very long stage of inflation with a generic potential. In addition we find superimposed oscillations, which would get imprinted on any kind of perturbations, scalars and tensors. Such oscillations have a period of 4–5 efolds and an amplitude which is typically less than a few percent and decreases linearly with f {sub γ}. We also stress that the curvature perturbation on uniform density slices should be sensitive to slow-roll parameters related to ρ{sub R} rather than φ-dot {sup 2}/2 and we discuss the existence of friction terms acting on the perturbations, although we postpone a calculation of the power spectrum and of non-gaussianity to future work and we simply define and compute suitable slow roll parameters. Finally we stress that this scenario may be realized in the axion case, if the coupling 1/ f {sub γ} to U(1) (photons) is much larger than the coupling 1/ f {sub G} to non-abelian gauge fields (gluons), since the latter sets the range of the potential and therefore the maximal allowed φ{sub 0∼} f {sub G}.

  19. Techniques for computing reactivity changes caused by fuel axial expansion in LMR's

    International Nuclear Information System (INIS)

    Khalil, H.

    1988-01-01

    An evaluation is made of the accuracy of methods used to compute reactivity changes caused by axial fuel relocation in fast reactors. Results are presented to demonstrate the validity of assumptions commonly made such as linearity of reactivity with fuel elongation, additivity of local reactivity contributions, and the adequacy of standard perturbation techniques. Accurate prediction of the reactivity loss caused by axial swelling of metallic fuel is shown to require proper representation of the burnup dependence of the expansion reactivity. Some accuracy limitations in the methods used in transient analyses, which are based on the use of fuel worth tables, are identified, and efficient ways to improve accuracy are described. Implementation of these corrections produced expansion reactivity estimates within 5% of higher-order method for a metal-fueled FFTF core representation. 18 refs., 3 figs., 3 tabs

  20. Calculation of effect of burnup history on spent fuel reactivity based on CASMO5

    International Nuclear Information System (INIS)

    Li Xiaobo; Xia Zhaodong; Zhu Qingfu

    2015-01-01

    Based on the burnup credit of actinides + fission products (APU-2) which are usually considered in spent fuel package, the effect of power density and operating history on k_∞ was studied. All the burnup calculations are based on the two-dimensional fuel assembly burnup program CASMO5. The results show that taking the core average power density of specified power plus a bounding margin of 0.0023 to k_∞, and taking the operating history of specified power without shutdown during cycle and between cycles plus a bounding margin of 0.0045 to k_∞ can meet the bounding principle of burnup credit. (authors)

  1. CHAR and BURNMAC - burnup modules of the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1986-03-01

    In the AUS neutronics code system, the burnup module CHAR solves the nuclide depletion equations by an analytic technique in a number of spatial zones. CHAR is usually used as one component of a lattice burnup calculation but contains features which also make it suitable for some global burnup calculations. BURNMAC is a simple accounting module based on the assumption that cross sections for a rector zone depend only on irradiation. BURNMAC is used as one component of a global calculation in which burnup is achieved by interpolation in the cross sections produced from a previous lattice calculation

  2. Triton burnup study using scintillating fiber detector on JT-60U

    International Nuclear Information System (INIS)

    Harano, Hideki

    1997-09-01

    The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV α particles behave. The α particles' behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton's behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector's directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R p -scan and n e -scan experiments have been performed. The R p -scan experiment indicates that the triton's transport was increased as the ripple amplitude over the triton became larger. In the n e -scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs

  3. Triton burnup study using scintillating fiber detector on JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Harano, Hideki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-09-01

    The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV {alpha} particles behave. The {alpha} particles` behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton`s behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector`s directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R{sub p}-scan and n{sub e}-scan experiments have been performed. The R{sub p}-scan experiment indicates that the triton`s transport was increased as the ripple amplitude over the triton became larger. In the n{sub e}-scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs.

  4. Calculation of isotope burn-up and change in efficiency of absorbing elements of WWER-1000 control and protection system during burn-up

    International Nuclear Information System (INIS)

    Timofeeva, O.A.; Kurakin, K.U.

    2006-01-01

    The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)

  5. Implementation of burnup in FERM nodal computer code

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Nakata, H.

    1986-01-01

    In this work a spatial burnup scheme and feedback effects has been implemented into the FERM [1] ('Finite Element Response Matrix') program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assemblywise calculation and pointwise calculation. The results have been compared with the results obtained by CITATION [2] program and showed that the processing time in the FERM code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt

  6. Burnup dependence of coolant void reactivity for ACR-1000 cell

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Hebert, A.; Roubstov, D.; Altiparmakov, D.; Irish, D.

    2008-01-01

    The Advanced Candu Reactor (ACR-1000) is light water cooled, fueled with enriched uranium and has a smaller lattice pitch than the Candu-6. As a result, the neutronic behavior of the ACR-1000 cell is expected to be somewhat different from that of the Candu-6 leading to a negative coolant void reactivity (CVR). Here we evaluate the CVR for the ACR-1000 cell using the lattice code DRAGON and compare our results with those obtained using the code WIMS-AECL. The differences observed between these two codes for the burnup dependence of the CVR is mainly explained in terms of the specific cell leakage model used by both codes. (authors)

  7. Development and verification of Monte Carlo burnup calculation system

    International Nuclear Information System (INIS)

    Ando, Yoshihira; Yoshioka, Kenichi; Mitsuhashi, Ishi; Sakurada, Koichi; Sakurai, Shungo

    2003-01-01

    Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)

  8. Burn-up TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

    Persic, A.; Ravnik, M.; Zagar, T.

    1998-01-01

    Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

  9. Preparation of computer codes for analyzing sensitivity coefficients of burnup characteristics (2) (Contract research, translated document)

    International Nuclear Information System (INIS)

    Hanaki, Hiroshi; Sanda, Toshio; Ohashi, Masahisa

    2008-10-01

    To develop nuclear design of LMFBR cores, they are important subjects of research and development to improve the accuracy in nuclear design of large LMFBR cores and to design highly efficient core more rationally. The adjusted nuclear cross-sections library has been made by being reflected the result of critical experiment of the JUPITER, etc. effectively as much as possible. And the distinct improvement of the accuracy in nuclear design of large LMFBR cores has been achieved. In the design of large LMFBR cores, however, it is important to accurately estimate not only nuclear characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. Therefore, it is thought to improve the prediction accuracy for burnup characteristics using many burnup data of 'Joyo' effectively. It is thought the best way to adjust cross sections using sensitivity coefficients of burnup characteristics to utilize burnup data of 'Joyo'. It is able to know the accuracy quantitatively for burnup characteristics of large LMFBR by analyzing the sensitivity coefficients. Therefore in this work computer codes for analyzing sensitivity coefficients of burnup characteristics had been prepared since 1992. In 1992 cross-section adjustment was done by using the data of 'Joyo' and the effect was studied. In this year the adequacy of the codes was studied with a view of applying of design of large LMFBR cores. The results are as follows: (1) The computer codes which could analyze sensitivity coefficients of burnup characteristics taking into consideration plural cycles and refueling were prepared, therefore it came of be able to adjust cross sections using burnup data and to estimate the accuracy for design of large LMFBR cores. The characteristics are not only burnup reactivity loss, breeding ratio but also number density, criticality, reactivity worth, reaction rate ratio, and reaction rate

  10. Transitions in axial morphology along the Southeast Indian Ridge

    Science.gov (United States)

    Ma, Ying; Cochran, James R.

    1996-07-01

    Shipboard bathymetric and magnetic profiles across the Southeast Indian Ridge (SEIR) were analyzed in order to examine the nature of along-axis variations in axial morphology at this intermediate spreading rate ridge. Three types of axial morphology are observed along the SEIR: an axial high, a shallow (200-700 m deep) axial valley and a deep (>1000 m deep) axial valley. An axial high is found to the east of the Australian-Antarctic Discordance (AAD) (east of 128°E) and between 82°E and 104°E. A shallow rift valley is found from 104°E to 114°E and from 82°E westward past the Amerstdam/St. Paul hotspot (ASP) to about 30°S, 75°E. Deep rift valleys are found from 114°E to 128°E in the vicinity of the AAD and from the Indian Ocean Triple Junction (IOTJ) at 25°S, 70°E to about 30°S, 75°E. The transition near 30°S occurs in an area of constant zero-age depth and does not appear to result from an increase in mantle temperature. It could be the result of the rapid increase in spreading rate along that portion of the SEIR. The most likely cause of the other transitions in axial morphology is variations in mantle temperature. The transitions between the different types of axial morphology are well defined and occur over a limited distance. Transitions in axial morphology are accompanied by significant changes in ridge flank topographic roughness. The transitions from axial valleys to axial highs are also accompanied by changes in the amplitude of the seafloor magnetic anomalies. Our observations suggest that there are distinct modes rather than a continuum of axial morphology on the SEIR and that there appears to be a "threshold" mechanism for a rapid change between different states of axial morphology. The ASP has only a limited influence on the SEIR. The ridge axis is marked by an axial valley for the entire distance from the IOTJ up to and past the ASP. The ridge axis becomes shallower as the ASP is approached from the northwest but only by about 300 m over

  11. Computer programs for TRIGA calibration, burnup evaluation, and bookkeeping

    International Nuclear Information System (INIS)

    Nelson, George W.

    1978-01-01

    Several computer programs have been developed at the University of Arizona to assist the direction and operation of the TRIGA Reactor Laboratory. The programs fall into the following three categories: 1. Programs for calculation of burnup of each fuel element in the reactor core, for maintaining an inventory of fuel element location and fissile content at any time, and for evaluation of the reactivity effects of burnup or proposed fuel element rearrangement in the core. 2. Programs for evaluation, function fitting, and tabulation of control rod measurements. 3. Bookkeeping programs to summarize and tabulate reactor runs and irradiations according to time, energy release, purpose, responsible party, etc. These summarized data are reported in an annual operating report for the facility. The use of these programs has saved innumerable hours of repetitious work, assuring more accurate, objective results, and requiring a minimum of effort to repeat calculations when input data are modified. The programs are written in FORTRAN-IV, and have been used on a CDC-6400 computer. (author)

  12. chemical determination of burnup ratio in nuclear fuels

    International Nuclear Information System (INIS)

    Guereli, L.

    1997-01-01

    Measurements of the extent of fission are important to determine the irradiation performance of a nuclear fuel. The energy released per unit mass of uranium (burnup) can be determined from measurement of the percent of heavy atoms that have fissioned during irradiation.The preferred method for this determination is choosing a suitable fission monitor (usually ''1''4''8Nd) and its determination after separation from the fuel matrix. In thermal reactor fuels where the only heavy element in the starting material is uranium, uranium depletion can be used for burnup determination. ''2''3''5U depletion method requires measurement of uranium isotopic ratios of both irradiated and unirradiated fuel. Isotopic ratios can be determined by thermal ionization mass spectrometer following separation of uranium from the fuel matrix. Separation procedures include solvent extraction, ion exchange and anion exchange chromatography. Another fission monitor used is ''1''3''9La determination by HPLC. Because La is monoisotopic (''1''3''9La) in the fuel, it can be determined by chemical analysis techniques

  13. Higher order methods for burnup calculations with Bateman solutions

    International Nuclear Information System (INIS)

    Isotalo, A.E.; Aarnio, P.A.

    2011-01-01

    Highlights: → Average microscopic reaction rates need to be estimated at each step. → Traditional predictor-corrector methods use zeroth and first order predictions. → Increasing predictor order greatly improves results. → Increasing corrector order does not improve results. - Abstract: A group of methods for burnup calculations solves the changes in material compositions by evaluating an explicit solution to the Bateman equations with constant microscopic reaction rates. This requires predicting representative averages for the one-group cross-sections and flux during each step, which is usually done using zeroth and first order predictions for their time development in a predictor-corrector calculation. In this paper we present the results of using linear, rather than constant, extrapolation on the predictor and quadratic, rather than linear, interpolation on the corrector. Both of these are done by using data from the previous step, and thus do not affect the stepwise running time. The methods were tested by implementing them into the reactor physics code Serpent and comparing the results from four test cases to accurate reference results obtained with very short steps. Linear extrapolation greatly improved results for thermal spectra and should be preferred over the constant one currently used in all Bateman solution based burnup calculations. The effects of using quadratic interpolation on the corrector were, on the other hand, predominantly negative, although not enough so to conclusively decide between the linear and quadratic variants.

  14. Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.

    1996-01-01

    The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)

  15. Application of reactivity method to MTR fuel burn-up measurement

    International Nuclear Information System (INIS)

    Zuniga, A.; Ravnik, M.; Cuya, R.

    2001-01-01

    Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

  16. Analysis on burn-up behaviors for accelerator-driven sub-critical facility

    International Nuclear Information System (INIS)

    Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qinbiao; Ding Dazhao

    2000-01-01

    An analysis is performed on burn-up behaviors for accelerator-driven sub-critical reactor by means of the code PASC-1 for neutronics calculation, the code CBURN for burn-up calculation and 44 group constants is processed by CENDL-2 and ENDF/B-6 using NJOY-91.91

  17. Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Rodriguez Rivada, A.

    2014-07-01

    Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)

  18. Determination of nuclear fuel burn-up using mass spectrometric techniques

    International Nuclear Information System (INIS)

    Saha, B.; Bagyalakshmi, R.; Periaswami, G.; Kavimandan, V.D.; Chitambar, S.A.; Jain, H.C.; Mathews, C.K.

    1977-01-01

    Determination of burn-up using a stable fission product monitor such as 148 Nd and heavy elements, determined by isotope dilution mass spectrometry gives the most accurate data. This report describes the work carried out to standardise the conditions for burn-up determination. Some typical results are given. (author)

  19. The application of burnup credit for spent fuel operations in the United Kingdom

    International Nuclear Information System (INIS)

    Bowden, R.

    1998-01-01

    This paper begins by outlining the structure of the nuclear industry in the United Kingdom. It then sets out the methodology of burnup credit, and provides a brief discussion of the validation and robustness of the calculational route. This leads to a description of both the current and intended applications of burnup credit in the United Kingdom. (author)

  20. Microstructural change and its influence on fission gas release in high burnup UO 2 fuel

    Science.gov (United States)

    Une, K.; Nogita, K.; Kashibe, S.; Imamura, M.

    1992-06-01

    The microstructural change of UO 2 fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO 2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO 2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO 2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

  1. The Width of High Burnup Structure in LWR UO2 Fuel

    International Nuclear Information System (INIS)

    Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

    2007-01-01

    The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

  2. Simulation of High Burnup Structure in UO2 Using Potts Model

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho

    2009-01-01

    The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels

  3. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

  4. Improvements on burnup chain model and group cross section library in the SRAC system

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.

    1992-01-01

    Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)

  5. A survey of previous and current industry-wide efforts regarding burnup credit

    International Nuclear Information System (INIS)

    Jones, R.H.

    1989-01-01

    Sandia has examined the matter of burnup credit from the perspective of physics, logistics, risk, and economics. A limited survey of the nuclear industry has been conducted to get a feeling for the actual application of burnup credit. Based on this survey, it can be concluded that the suppliers of spent fuel storage and transport casks are in general agreement that burnup credit offers the potential for improvements in cask efficiency without increasing the risk of accidental criticality. The actual improvement is design-specific but limited applications have demonstrated that capacity increases in the neighborhood of 20 percent are not unrealistic. A number of these vendors acknowledge that burnup credit has not been reduced to practice in cask applications and suggest that operational considerations may be more important to regulatory acceptance than to the physics. Nevertheless, the importance of burnup credit to the nuclear industry as a cask design and analysis tool has been confirmed by this survey

  6. Regulatory status of burnup credit for storage and transport of spent fuel in Germany

    International Nuclear Information System (INIS)

    Neuber, J.C.; Schweer, H.H.; Johann, H.G.

    2001-01-01

    This paper describes the regulatory status of burnup credit applications to pond storage and dry-cask transport and storage of spent fuel in Germany. Burnup credit for wet storage of LWR fuel at nuclear power plants has to comply with the newly developed safety standard DIN 25471. This standard establishes the safety requirements for burnup credit criticality safety analysis of LWR fuel storage ponds and gives guidance on meeting these requirements. Licensing evaluations of dry transport systems are based on the application of the IAEA Safety Standards Series No.ST-1. However, because of the fact that burnup credit for dry-cask transport becomes more and more inevitable due to increasing initial enrichment of the fuel, and because of the increasing importance of dry-cask storage in Germany, the necessity of giving regulatory guidance on applying burnup credit to dry-cask transport and storage is seen. (author)

  7. Assessment of US NRC fuel rod behavior codes to extended burnup

    International Nuclear Information System (INIS)

    Laats, E.T.; Croucher, D.W.; Haggag, F.M.

    1982-01-01

    The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available

  8. Cell verification of parallel burnup calculation program MCBMPI based on MPI

    International Nuclear Information System (INIS)

    Yang Wankui; Liu Yaoguang; Ma Jimin; Wang Guanbo; Yang Xin; She Ding

    2014-01-01

    The parallel burnup calculation program MCBMPI was developed. The program was modularized. The parallel MCNP5 program MCNP5MPI was employed as neutron transport calculation module. And a composite of three solution methods was used to solve burnup equation, i.e. matrix exponential technique, TTA analytical solution, and Gauss Seidel iteration. MPI parallel zone decomposition strategy was concluded in the program. The program system only consists of MCNP5MPI and burnup subroutine. The latter achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, and data exchanging with MCNP5MPI. Also, the program was verified with the pressurized water reactor (PWR) cell burnup benchmark. The results show that it,s capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

  9. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  10. Application of burnup credit in spent fuel management at Russian NPPs

    International Nuclear Information System (INIS)

    Koulikov, V.I.; Makarchuk, T.F.; Tikhonov, N.S.

    1998-01-01

    The article concerns implementation of burnup credit in spent fuel storage and transportation. Some of the problems with increased enrichment fuel can be resolved by use of modified transport methodology. Such as shipping in gas-filled casks only, reduced number of assemblies in casks, etc. However, the use of modified schemes of transportation results in essential financial losses. An actinide-only burnup credit is taken into account in most part of criticality calculations, and a parameter limiting loading of spent fuel in the cask or the repository is the avenge value of burnup on an assembly. The main method of burnup depth definition is its defect measurement. A short description of devices for measurement as well as some technical results of suing burnup credit approach in storage and transport are given. (author)

  11. Determination of burn-up of irradiated nuclear fuels using mass spectrometry

    International Nuclear Information System (INIS)

    Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

    2017-01-01

    Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

  12. Development of burnup methods and capabilities in Monte Carlo code RMC

    International Nuclear Information System (INIS)

    She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

    2013-01-01

    Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

  13. Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

    International Nuclear Information System (INIS)

    Wagner, John C.; Parks, Cecil V.; Mueller, Don; Gauld, Ian C.

    2010-01-01

    Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

  14. Fission-gas release in fuel performing to extended burnups in Ontario Hydro nuclear generating stations

    International Nuclear Information System (INIS)

    Floyd, M.R.; Novak, J.; Truant, P.T.

    1992-06-01

    The average discharge burnup of CANDU fuel is about 200 MWh/kgU. A significant number of 37-element bundles have achieved burnups in excess of 400 MWh/kgU. Some of these bundles have experienced failures related to their extended operation. To date, hot-cell examinations have been performed on fuel elements from nine 37-element bundles irradiated in Bruce NGS-A that have burnups in the range of 300-800 MWh/kgU. 1 Most of these have declining power histories from peak powers of up to 59 kW/m. Fission-gas releases of up to 26% have been observed and exhibit a strong dependence on fuel power. This obscures any dependence on burnup. The extent of fission-gas release at extended burnups was not predicted by low-burnup code extrapolations. This is attributed primarily to a reduction in fuel thermal conductivity which results in elevated operating temperatures. Reduced conductivity is due, at least in part, to the buildup of fission products in the fuel matrix. Some evidence of hyperstoichiometry exists, although this needs to be further investigated along with any possible relation to CANLUB graphite coating behaviour and sheath oxidation. Residual tensile sheath strains of up to 2% have been observed and can be correlated with fuel power/fission-gas release. SCC 2 -related defects have been observed in the sheath and endcaps of elements from bundles experiencing declining power histories to burnups in excess of 500 MWh/kgU. This indicates that the current recommended burnup limit of 450 MWh/kgU is justified. SCC-related defects have also been observed in ramped bundles having burnups < 450 MWh/kgU. Hence, additional guidelines are in place for power ramping extended-burnup fuel

  15. Thermocapillary Convection in Floating Zone with Axial Magnetic Fields

    Science.gov (United States)

    Liang, Ruquan; Yang, Shuo; Li, Jizhao

    2014-02-01

    Numerical simulations on the effects of axial magnetic fields on the thermocapillary convection in a liquid bridge of silicone-oil-based ferrofluid under zero gravity have been conducted. The Navier-Stokes equations coupled with the energy conservation equation are solved on a staggered grid, and the mass conserving level set approach is used to capture the free surface deformation of the liquid bridge. The obvious effects of the magnetic fields on the flow pattern as well as the velocity and temperature distributions in the liquid bridge can be detected. The axial magnetic fields suppress the thermocapillary convection and a stagnant flow zone is formed between the circulating flow and the symmetric axis as the magnetic fields increase. The axial magnetic fields affect not only the velocity level inside the liquid bridge but also the velocity level on the free surface. The temperature contours near the free surface illustrates conduction-type temperature profiles at moderate strength fields.

  16. New Burnup Calculation System for Fusion-Fission Hybrid System

    International Nuclear Information System (INIS)

    Isao Murata; Shoichi Shido; Masayuki Matsunaka; Keitaro Kondo; Hiroyuki Miyamaru

    2006-01-01

    Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise

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

    International Nuclear Information System (INIS)

    Hanson, B.D.

    1998-07-01

    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, UO 2 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 UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 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 UO 2 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 UO 2 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. A note on axial symmetries

    International Nuclear Information System (INIS)

    Beetle, Christopher; Wilder, Shawn

    2015-01-01

    This note describes how to characterize and normalize an axial Killing field on a general Riemannian geometry or four-dimensional Lorentzian geometry. No global assumptions are necessary, such as that the orbits of the Killing field all have period 2π. Rather, any Killing field that vanishes at at least one point necessarily has the expected global properties. (note)

  19. Axial structure of the nucleon

    Energy Technology Data Exchange (ETDEWEB)

    Veronique Bernard; Latifa Elouadrhiri; Ulf-G Meissner

    2002-01-01

    We review the current status of experimental and theoretical understanding of the axial nucleon structure at low and moderate energies. Topics considered include (quasi)elastic (anti)neutrino-nucleon scattering, charged pion electroproduction off nucleons and ordinary as well as radiative muon capture on the proton.

  20. Investigation of axial power gradients near a control rod tip

    Energy Technology Data Exchange (ETDEWEB)

    Loberg, John, E-mail: John.Loberg@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Osterlund, Michael, E-mail: Michael.Osterlund@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Bejmer, Klaes-Hakan, E-mail: Klaes-Hakan.Bejmer@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Blomgren, Jan, E-mail: Jan.Blomgren@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Kierkegaard, Jesper, E-mail: Jesper.Kierkegaar@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden)

    2011-07-15

    Highlights: > Pin power gradients near BWR control rod tips have been investigated. > A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. > Small nodes increases pin power gradients; standard nodes underestimates gradients. > The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, {approx}15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

  1. Investigation of axial power gradients near a control rod tip

    International Nuclear Information System (INIS)

    Loberg, John; Osterlund, Michael; Bejmer, Klaes-Hakan; Blomgren, Jan; Kierkegaard, Jesper

    2011-01-01

    Highlights: → Pin power gradients near BWR control rod tips have been investigated. → A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. → Small nodes increases pin power gradients; standard nodes underestimates gradients. → The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, ∼15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

  2. Simulation of triton burn-up in JET plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, M J; Balet, B; Jarvis, O N; Stubberfield, P M [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

  3. 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).

  4. Visualization of fuel rod burnup analysis by Scilab

    International Nuclear Information System (INIS)

    Tsai, Chiung-Wen

    2013-01-01

    The goal of this technical note is to provide an alternative, the freeware Scilab, by which means we may construct custom GUIs and distribute them without extra constrains and cost. A post-processor has been constructed by Scilab to visualize the fuel rod burnup analysis data calculated by FRAPCON-3.4. This post-processor incorporates a graphical user interface (GUI), providing users a rapid overview of the characteristics of the numerical results with 2-D and 3-D graphs, as well as the animations of fuel temperature distribution. An assessment case input file provided by FRAPCON user group was applied to demonstrate the construction of a post-processor with GUI by object-oriented GUI tool, as well as the capability of visualization functions of Scilab

  5. Visualization of fuel rod burnup analysis by Scilab

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, Chiung-Wen, E-mail: d937121@oz.nthu.edu.tw

    2013-12-15

    The goal of this technical note is to provide an alternative, the freeware Scilab, by which means we may construct custom GUIs and distribute them without extra constrains and cost. A post-processor has been constructed by Scilab to visualize the fuel rod burnup analysis data calculated by FRAPCON-3.4. This post-processor incorporates a graphical user interface (GUI), providing users a rapid overview of the characteristics of the numerical results with 2-D and 3-D graphs, as well as the animations of fuel temperature distribution. An assessment case input file provided by FRAPCON user group was applied to demonstrate the construction of a post-processor with GUI by object-oriented GUI tool, as well as the capability of visualization functions of Scilab.

  6. User's manual for the reactor burnup system, REBUS

    International Nuclear Information System (INIS)

    Olson, A.P.; Regis, J.P.; Meneley, D.A.; Hoover, L.J.

    1972-01-01

    A user's manual for the REBUS System (REactor BUrnup System) is presented. Its primary purpose is to provide sufficient information about the REBUS capability to the user to ensure its efficient utilization. The current REBUS System either solves for the infinite time (equilibrium) operating conditions of a recycle system under fixed conditions, or solves for operating conditions during a single time step (non-equilibrium). The capability of studying various in-reactor fuel management and ex-reactor fuel management schemes has been included. REBUS has been operated with one- and two-dimensional diffusion theory neutronics solutions up to the present time. The model was specifically designed for extension to other neutronics models such as three-dimensional diffusion or transport theory and direct or synthesis solutions

  7. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems

    Directory of Open Access Journals (Sweden)

    Thomas Frosio

    2017-01-01

    Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.

  8. Validating analysis methodologies used in burnup credit criticality calculations

    International Nuclear Information System (INIS)

    Brady, M.C.; Napolitano, D.G.

    1992-01-01

    The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

  9. Oxide thickness measurement for monitoring fuel performance at high burnup

    International Nuclear Information System (INIS)

    Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

    1991-01-01

    For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

  10. Technical description of the burn-up software system MOP

    International Nuclear Information System (INIS)

    Schutte, C.K.

    1991-05-01

    The burn-up software system MOP is a research tool primary intended to study the behaviour of fission products in any reactor composition. Input data are multi-group cross-sections and data concerning the nuclide chains. An option is available to calculate a fundamental mode neutron spectrum for the specified reactor composition. A separate program can test the consistency of the specified nuclide chains. Options are available to calculate time-dependent cross-sections of lumped fission products and to take account of the leakage of gaseous fission products from the reactor core. The system is written in FORTRAN77 for a CYBER computer, using the operating system NOS/BE. The report gives a detailed technical description of the applied algorithms and the flow and storage of data. Information is provided for adapting the system to other computer configurations. (author). 5 refs.; 11 figs

  11. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    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. Approach to lithium burn-up effect in lithium ceramics

    International Nuclear Information System (INIS)

    Rasneur, B.

    1994-01-01

    The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

  13. Time resolved measurements of triton burnup in JET plasmas

    International Nuclear Information System (INIS)

    Conroy, S.; Jarvis, O.N.; Sadler, G.; Huxtable, G.B.

    1988-01-01

    Triton production from one branch of the deuteron-deuteron fusion reaction is routinely measured at 6 ms time intervals in JET plasma discharges by recording the 2.5 MeV neutrons produced in the other branch using a set of calibrated fission chambers. The burnup of the tritons is measured by detecting the 14 MeV t-d neutrons with a 0.2 cm 3 Si(Li) diode. The 2.5 MeV neutron flux can be used in a simple time dependent calculation based on classical slowing-down theory to predict the 14 MeV neutron flux. The measured flux and the triton slowing-down time are systematically lower than the values estimated from the key plasma parameters but the differences are within the experimental errors. (author). 19 refs, 8 figs

  14. Monte Carlo burnup codes acceleration using the correlated sampling method

    International Nuclear Information System (INIS)

    Dieudonne, C.

    2013-01-01

    For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step. In this document we present an original methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time we develop a theoretical model to study the features of the correlated sampling method to understand its effects on depletion calculations. In a third time the implementation of this method in the TRIPOLI-4 code will be discussed, as well as the precise calculation scheme used to bring important speed-up of the depletion calculation. We will begin to validate and optimize the perturbed depletion scheme with the calculation of a REP-like fuel cell depletion. Then this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes. (author) [fr

  15. Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach

    International Nuclear Information System (INIS)

    Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou

    2005-01-01

    A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)

  16. Full Core Burn-up Calculation at JRR-3 with MVP-BURN

    International Nuclear Information System (INIS)

    Komeda, Masao; Yamamoto, Kazuyoshi; Kusunoki, Tsuyoshi

    2008-01-01

    Research reactors use a burnable poison to suppress an excess reactivity in the beginning of reactor lifetime. The JRR-3 (Japan Research Reactor No.3) has used cadmium wires of radius 0.02 cm as a burnable poison. This report describes burn-up calculations of plate fuel models and full core models with MVP-BURN, which is a burn-up calculation code using Monte Carlo method and has been developed in JAEA (Japan Atomic Energy Agency). As the results of calculations of plate models, between a model composed of one burn-up region along the radius direction and a model composed of a few burn-up regions along the radius direction, the effective absorption cross section of 113 Cd has had different tendency on reaching approximate 40. day (10000 MWd/t). And as results of calculations of full core model, it has been indicated that k eff is almost same till approximate 80. day (22000 MWd/t) between a model composed of one burn-up region along the vertical direction and a model composed of a few burn-up regions along the vertical direction. However difference of 113 Cd burn-up becomes pronounced and each k eff makes a difference after 80. day. (authors)

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

    International Nuclear Information System (INIS)

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-01-01

    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

  18. Analyzing the BWR rod drop accident in high-burnup cores

    International Nuclear Information System (INIS)

    Diamond, D.J.; Neymotin, L.; Kohut, P.

    1995-01-01

    This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ''rim'' effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions

  19. Determination of enrichment of recycle uranium fuels for different burnup values

    International Nuclear Information System (INIS)

    Zabunoglu, Okan H.

    2008-01-01

    Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000-50,000 MWd/tonU

  20. Development and Applications of a Prototypic SCALE Control Module for Automated Burnup Credit Analysis

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2001-01-01

    Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale burnup credit analyses that can be performed in a reasonable amount of time. STARBUCS is a new prototypic analysis sequence being developed for the SCALE code system to perform automated criticality calculations of spent fuel systems employing burnup credit. STARBUCS is designed to help analyze the dominant burnup credit phenomena including spatial burnup gradients and isotopic uncertainties. A search capability also allows STARBUCS to iterate to determine the spent fuel parameters (e.g., enrichment and burnup combinations) that result in a desired k eff for a storage configuration. Although STARBUCS was developed to address the analysis needs for spent fuel transport and storage systems, it provides sufficient flexibility to allow virtually any configuration of spent fuel to be analyzed, such as storage pools and reprocessing operations. STARBUCS has been used extensively at Oak Ridge National Laboratory (ORNL) to study burnup credit phenomena in support of the NRC Research program

  1. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  2. Burnup effect on nuclear fuel cycle cost using an equilibrium model

    International Nuclear Information System (INIS)

    Youn, S. R.; Kim, S. K.; Ko, W. I.

    2014-01-01

    The degree of fuel burnup is an important technical parameter to the nuclear fuel cycle, being sensitive and progressive to reduce the total volume of process flow materials and eventually cut the nuclear fuel cycle costs. This paper performed the sensitivity analysis of the total nuclear fuel cycle costs to changes in the technical parameter by varying the degree of burnups in each of the three nuclear fuel cycles using an equilibrium model. Important as burnup does, burnup effect was used among the cost drivers of fuel cycle, as the technical parameter. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once Through Cycle(PWR-OT), PWR-MOX Recycle, Pyro-SFR Recycle. These fuel cycles are most likely to be adopted in the foreseeable future. As a result of the sensitivity analysis on burnup effect of each three different nuclear fuel cycle costs, PWR-MOX turned out to be the most influenced by burnup changes. Next to PWR-MOX cycle, in the order of Pyro-SFR and PWR-OT cycle turned out to be influenced by the degree of burnup. In conclusion, the degree of burnup in the three nuclear fuel cycles can act as the controlling driver of nuclear fuel cycle costs due to a reduction in the volume of spent fuel leading better availability and capacity factors. However, the equilibrium model used in this paper has a limit that time-dependent material flow and cost calculation is impossible. Hence, comparative analysis of the results calculated by dynamic model hereafter and the calculation results using an equilibrium model should be proceed. Moving forward to the foreseeable future with increasing burnups, further studies regarding alternative material of high corrosion resistance fuel cladding for the overall

  3. Determination of the burn-up in fuels of the MTR type by means of gamma spectroscopy with crystal of INa(Tl)

    International Nuclear Information System (INIS)

    Kestelman, A.J.

    1988-01-01

    One of the responsibilities of the Laboratory of Analysis by Neutronic Activation of the RA-6 reactor is to determine the burn-up in fuels of the MTR type. In order to gain experience, up to the arrival of the hyperpure Germanium detector (HPGe) to be used in normal operation, preliminary measurements with a crystal of INa(Tl) were made. The fuel elements used are originated in the RA-3 reactor, with a decay superior to the thirteen years. For this reason, the unique visible photoelectric peak is the one of Cs-137, owing to the low resolution of the INa(Tl). After preliminary measurements, the profiles of burn-up, rectified by attenuation, were measured. Once the efficiency of the detector was determined, the calculation of the burn-up was made; for the element No. 144, a value of 21.6 ± 2.9 g was obtained to be compared with the value 21.9 g which was the evaluation made by the operators. (Author) [es

  4. Construction and tests of a gamma device for experimental measurements of burnup of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Brandao Junior, F.A.

    1982-01-01

    The gamma-scanning method is an important tool for the measurement of burnup of nuclear reactor fuel. The adequate knowledge of burnup allows for a better inventory of 'sensitive' fissile materials, better fuel management and provides insight on fuel behaviour and safety margins. This paper is related to the description, construction and operation of a first gamma scanning device, tested by irradiation of prototype PWR fuel pins, 14 cm long, in a Triga Mark-I reactor at very low power. Despite the limitations imposed by the low burnup, the experiment permitted a good checking of the main physical concepts and devices involved in the method. (Author) [pt

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

    International Nuclear Information System (INIS)

    Manolova, M.

    1998-01-01

    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)

  6. SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors

    International Nuclear Information System (INIS)

    Tsuchihashi, K.; Ishiguro, Y.; Kaneko, K.; Ido, M.

    2004-01-01

    1 - Description of program or function: General neutronics calculation including cell calculation with burn-up, core calculation for any type of thermal reactor. Core burn-up calculation and fuel management by an auxiliary code. 2 - Method of solution: Collision probability method, 1D and 2D Sn for cell calculation; 1D, 2D and 3D diffusion for core calculation. 3 - Restrictions on the complexity of the problem: 20 regions for a continuous energy resonance absorption calculation and 16 steps for cell burn-up

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

    International Nuclear Information System (INIS)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro

    2011-01-01

    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 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)

  8. Development of continuous energy Monte Carlo burn-up calculation code MVP-BURN

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Nakagawa, Masayuki; Sasaki, Makoto

    2001-01-01

    Burn-up calculations based on the continuous energy Monte Carlo method became possible by development of MVP-BURN. To confirm the reliably of MVP-BURN, it was applied to the two numerical benchmark problems; cell burn-up calculations for High Conversion LWR lattice and BWR lattice with burnable poison rods. Major burn-up parameters have shown good agreements with the results obtained by a deterministic code (SRAC95). Furthermore, spent fuel composition calculated by MVP-BURN was compared with measured one. Atomic number densities of major actinides at 34 GWd/t could be predicted within 10% accuracy. (author)

  9. Discharge Burnup Evaluation of Natural Uranium Loaded CANFLEX-43 Fuel Bundle

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Kim, Yong Hee; Kim, Won Young; Park, Joo Hwan

    2009-11-01

    Using WIMS-AECL code, which is 2-dimensional lattice core used in CANDU physics calculation, the discharge burnup of the natural uranium loaded CANFLEX-43 fuel bundle was evaluated by comparing the discharge burnup of standard 37 element fuel bundle. When the discharge burnup of the standard 37 element fuel is 7,200 MWd/MTU, that of the CANFLEX 43 fuel bundle was evaluated as 7,077 MWd/MTU, by applying the same lattice conditions for both fuel bundles

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

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    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

  11. 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)

  12. Approach for implementing burnup credit in high-capacity truck casks

    International Nuclear Information System (INIS)

    Boshoven, J.; Hopf, J.; Su, S.

    1991-01-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks in 1992. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for the higher enrichments. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA presents here an approach to burnup credit analysis to be included in the Safety Analysis Report for Packaging (SARP). 6 refs., 2 figs., 5 tabs

  13. High Burnup Fuel: Implications and Operational Experience. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2016-08-01

    This publication reports on the outcome of a technical meeting on high burnup fuel experience and economics, held in Buenos Aires, Argentina in 2013. The purpose of the meeting was to revisit and update the current operational experience and economic conditions associated with high burnup fuel. International experts with significant experience in experimental programmes on high burnup fuel discussed and evaluated physical limitations at pellet, cladding and structural component levels, with a wide focus including fabrication, core behaviour, transport and intermediate storage for most types of commercial nuclear power plants

  14. Criterion for burn-up conditions in gas-cooled cryogenic current leads

    International Nuclear Information System (INIS)

    Bejan, A.; Cluss, E.M. Jr.

    1976-01-01

    Superconducting magnets are energized through helium vapour-cooled cryogenic current leads operating at high ratios of current to mass flow. The high current operation where lead temperature, runaway, and eventual burn-up are likely to occur is investigated. A simple criterion for estimating the burn-up operation conditions (current, mass flow) for a given lead geometry (cross-sectional area, length, heat exchanger area) is presented. This article stresses the role played by the available heat exchanger area in avoiding burn-up at high ratios of current to mass flow. (author)

  15. The research on burnup characteristic of doping burnable poison in PWR

    International Nuclear Information System (INIS)

    Qiang Shenglong; Qin Dong; Chai Xiaoming; Yao Dong

    2014-01-01

    In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons (such as natural element, manual nuclide and soluble boron) in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf, Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core. (authors)

  16. Axial magnetic field produced by axially and radially magnetized permanent rings

    International Nuclear Information System (INIS)

    Peng, Q.L.; McMurry, S.M.; Coey, J.M.D.

    2004-01-01

    Axial magnetic fields produced by axially and radially magnetized permanent magnet rings were studied. First, the axial magnetic field produced by a current loop is introduced, from which the axial field generated by an infinitely thin solenoid and by an infinitely thin current disk can be derived. Then the axial fields produced by axially and by radially magnetized permanent magnet rings can be obtained. An analytic formula for the axial fields produced by two axially magnetized rings is given. A permanent magnet with a high axial gradient field is fabricated, the measured results agree with the theoretical calculation very well. As an example, the axial periodic field produced by an arrangement of alternating axially and radially magnetized rings has been discussed

  17. View of the Axial Field Spectrometer

    CERN Multimedia

    1980-01-01

    The Axial Field Spectrometer, with the vertical uranium/scintillator calorimeter and the central drift chamber retracted for service. One coil of the Open Axial Field Magnet is just visible to the right.

  18. Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  19. Axial-Centrifugal Compressor Program

    Science.gov (United States)

    1975-10-01

    Assembly . .. . .... ..... 33 5 Tie Bolt...... .. .. .. .. . *.. .. .. .. .. .. ... 34 6 Axial Compressor Rotor Assembly Runouts . . .. . 34 7 CCV Blow...1.796 Impeller Slip Factor ’Ce2/U 2 ) .91 Impeller Wheel Speed ft/sec 1992.2 Impellet ’.ip Radius in. 3.780 Blade Tip Metal Angle- deg 0 Numbec of Blades...test item to the next Phase V component test. The test vehicle final balance levels and rotor runouts were normal at teardown, and no rubsI were

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  1. Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

    International Nuclear Information System (INIS)

    Pinem, Surian; Liem, Peng Hong; Sembiring, Tagor Malem; Surbakti, Tukiran

    2016-01-01

    Highlights: • Burnup measurement of fuel elements comprising the new equilibrium LEU silicide core of RSG GAS. • The burnup measurement method is based on a linear relationship between reactivity and burnup. • Burnup verification was conducted using an in-house, in-core fuel management code BATAN-FUEL. • A good agreement between the measured and calculated burnup was confirmed. • The new fuel management strategy was confirmed and validated. - Abstract: After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

  2. Synthesis of Axial Power Distribution Using 5-Level Ex-core Detector in a Core Protection System

    International Nuclear Information System (INIS)

    Koo, Bon-Seung; Lee, Chung-Chan; Zee, Sung-Quun

    2007-01-01

    In ABB-CE digital plants, Core Protection Calculator System (CPCS) is used for a core protection based on several online measured system parameters including 3- level safety grade ex-core detector signals. The CPCS provides four independent channels for the departure from a nucleate boiling ratio (DNBR) and local power density (LPD) trip signals to the reactor protection system. Each channel consists of a core protection calculator (CPC) and a control element assembly calculator (CEAC). The cubic spline synthesis technique has been used in online calculations of the core axial power distributions using 3-level ex-core detector signals in CPC. The pre-determined cubic spline function sets are used depending on the characteristics of the ex-core detector responses. But this method shows large power distribution errors for the extremely skewed axial shapes due to restrictive function sets and an incorrect SAM value. Especially thus situation is worse at a higher burnup. To solve these problems, the cubic spline function sets are improved and it is demonstrated that the axial power shapes can be synthesized more accurately with the new function sets than those of a conventional CPC. In this paper, synthesis of an axial power distribution using a 5-level ex-core detector is described and the axial power distributions are compared between 3-level and 5-level ex-core detector systems

  3. Axial channeling of boron ions into silicon

    International Nuclear Information System (INIS)

    La Ferla, A.; Galvagno, G.; Raineri, V.; Setola, R.; Rimini, E.; Carnera, A.; Gasparotto, A.

    1992-01-01

    Channeling boron implants were performed into (100) and (110) silicon substrates in the energy range 80-700 keV. The dose ranged between 3.5x10 11 and 1x10 15 atoms/cm 2 . The axial channeling concentration profiles of implanted B + were compared with that obtained for incidence along the random direction of the crystal and with that obtained by implantation in amorphous silicon. The electrical and chemical boron distributions were obtained by spreading resistance and secondary ion mass spectrometry measurements, respectively. The inelastic stopping power, S c , was extracted from the experimental maximum ranges for the [100] and [110] axis. The energy dependence of the electronic stopping power is given by S e = KE p with p [100] = 0.469±0.010 and p [110] = 0.554±0.004. Simulations obtained by the MARLOWE code, using the Oen-Robinson impact parameter dependent formula, for the electronic energy loss reproduce quite well the experimental depth profiles. (orig.)

  4. Development and benchmark verification of a parallelized Monte Carlo burnup calculation program MCBMPI

    International Nuclear Information System (INIS)

    Yang Wankui; Liu Yaoguang; Ma Jimin; Yang Xin; Wang Guanbo

    2014-01-01

    MCBMPI, a parallelized burnup calculation program, was developed. The program is modularized. Neutron transport calculation module employs the parallelized MCNP5 program MCNP5MPI, and burnup calculation module employs ORIGEN2, with the MPI parallel zone decomposition strategy. The program system only consists of MCNP5MPI and an interface subroutine. The interface subroutine achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, data exchanging with MCNP5MPI. Also, the program was verified with the Pressurized Water Reactor (PWR) cell burnup benchmark, the results showed that it's capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

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

    International Nuclear Information System (INIS)

    1993-04-01

    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

  6. Conservatism in the actinide-only burnup credit for PWR spent nuclear fuel packages

    International Nuclear Information System (INIS)

    Lancaster, D.B.; Rahimi, M.; Thornton, J.

    1996-01-01

    In May 1995, the U.S. Department of Energy (DOE) submitted a topical report to the U.S. Nuclear Regulatory Commission (NRC) to gain actinide-only burnup credit for spent nuclear fuel (SNF) storage, transportation, or disposal packages. After approval of this topical report, DOE intends further submittals to the NRC to acquire additional burnup credit (e.g., the topical does not use fission products and is limited to only the first 100 yr of disposal). The NRC has responded to the topical with its preliminary questions. To aid in evaluation of the method, a review of the conservatism in the actinide-only burnup credit methodology was performed. An overview of the actinide-only burnup credit methodology is presented followed by a summary of the conservatism

  7. Proceedings of a workshop on the use of burnup credit in spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1989-10-01

    The Department of Energy sponsored a workshop on the use of burnup credit in the criticality design of spent fuel shipping casks on February 21 and 22, 1988. Twenty-five different presentations on many related topics were conducted, including the effects of burnup credit on the design and operation of spent fuel storage pools, casks and modules, and shipping casks; analysis and physics issues related to burnup credit; regulatory issues and criticality safety; economic incentives and risks associated with burnup credit; and methods for verifying spent fuel characteristics. An abbreviated version of the DOE workshop was repeated as a special session at the November 1988 American Nuclear Society Meeting in Washington, DC. Each of the invited speakers prepared detailed papers on his or her respective topic. The individual papers have been cataloged separately

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

    International Nuclear Information System (INIS)

    Yoshioka, Ken-ichi; Ando, Y.; Kumanomido, H.; Sasaki, T.; Mitsuhashi, I.; Ueda, M.

    2001-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-08-01

    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.

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

    International Nuclear Information System (INIS)

    2001-08-01

    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

  11. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

  12. Calculation of heat rating and burn-up for test fuel pins irradiated in DR 3

    International Nuclear Information System (INIS)

    Bagger, C.; Carlsen, H.; Hansen, K.

    1980-01-01

    A summary of the DR 3 reactor and HP1 rig design is given followed by a detailed description of the calculation procedure for obtaining linear heat rating and burn-up values of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially regarding features like end pellet contribution to power as a function of burn-up, gamma heat contributions, and evaluation of local values of heat rating and burn-up. Included in the report is also a description of the fast flux- and cladding temperature calculation techniques currently used. A good agreement between measured and calculated local burn-up values is found. This gives confidence to the detailed treatment of the data. (author)

  13. Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.

    1983-01-01

    A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

  14. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    International Nuclear Information System (INIS)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-01-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  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. The build-up and characterization of nuclear burn-up wave in a fast ...

    Indian Academy of Sciences (India)

    K V Anoop

    2018-02-07

    Feb 7, 2018 ... evaluating the quality of the wave by the researchers working in the field of nuclear burn-up wave build-up and propagation. Keywords. ... However, there are concerns relating to the nuclear safety, ... Simulation studies have.

  17. Technical Issues in the development of high burnup and long cycle fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  18. Technical Issues in the development of high burnup and long cycle fuel pellets

    International Nuclear Information System (INIS)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui

    2012-01-01

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  19. Calculation of pellet radial power distributions with a Monte Carlo burnup code

    International Nuclear Information System (INIS)

    Suzuki, Motomu; Yamamoto, Toru; Nakata, Tetsuo

    2010-01-01

    The Japan Nuclear Energy Safety Organization (JNES) has been working on an irradiation test program of high-burnup MOX fuel at Halden Boiling Water Reactor (HBWR). MOX and UO 2 fuel rods had been irradiated up to about 64 GWd/t (rod avg.) as a Japanese utilities research program (1st phase), and using those fuel rods, in-situ measurement of fuel pellet centerline temperature was done during the 2nd phase of irradiation as the JNES test program. As part of analysis of the temperature data, power distributions in a pellet radial direction were analyzed by using a Monte Carlo burnup code MVP-BURN. In addition, the calculated results of deterministic burnup codes SRAC and PLUTON for the same problem were compared with those of MVP-BURN to evaluate their accuracy. Burnup calculations with an assembly model were performed by using MVP-BURN and those with a pin cell model by using SRAC and PLUTON. The cell pitch and, therefore, fuel to moderator ratio in the pin cell calculation was determined from the comparison of neutron energy spectra with those of MVP-BURN. The fuel pellet radial distributions of burnup and fission reaction rates at the end of the 1st phase irradiation were compared between the three codes. The MVP-BURN calculation results show a large peaking in the burnup and fission rates in the pellet outer region for the UO 2 and MOX pellets. The SRAC calculations give very close results to those of the MVP-BURN. On the other hand, the PLUTON calculations show larger burnup for the UO 2 and lower burnup for the MOX pellets in the pellet outer region than those of MVP-BURN, which lead to larger fission rates for the UO 2 and lower fission rates for the MOX pellets, respectively. (author)

  20. Calculation study of the WWER-440 fuel performance for extended burnup

    International Nuclear Information System (INIS)

    Kujal, J.; Pazdera, F.; Barta, O.

    1984-01-01

    The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

  1. BEAVRS full core burnup calculation in hot full power condition by RMC code

    International Nuclear Information System (INIS)

    Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan

    2017-01-01

    Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.

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

    International Nuclear Information System (INIS)

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-01-01

    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 UO 2 matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus

  3. Calculational prediction of fuel burn-up for the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Phuoc Lan; Do Quang Binh

    2016-01-01

    In this paper, the method of expanding operators and functions in the neutron diffusion equations as chains of time variable is used for calculation of fuel burn-up of the Dalat nuclear reactors. A computer code, named BURREF, programmed in language Fortran-77 running on IBM PC-AT, has been developed based on this method to predict the fuel burn-up of the Dalat reactor. Some results will be presented here. (author)

  4. Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Okubo, Tsutomu

    2011-01-01

    Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

  5. Review of high burn-up RIA and LOCA database and criteria

    International Nuclear Information System (INIS)

    Vitanza, C.; Hrehor, M.

    2006-01-01

    This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

  6. A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyoungju; Park, Kwangheon; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of)

    2015-05-15

    The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years.

  7. Numerical solution of matrix exponential in burn-up equation using mini-max polynomial approximation

    International Nuclear Information System (INIS)

    Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

    2015-01-01

    Highlights: • We propose a new numerical solution of matrix exponential in burn-up depletion calculations. • The depletion calculation with extremely short half-lived nuclides can be done numerically stable with this method. • The computational time is shorter than the other conventional methods. - Abstract: Nuclear fuel burn-up depletion calculations are essential to compute the nuclear fuel composition transition. In the burn-up calculations, the matrix exponential method has been widely used. In the present paper, we propose a new numerical solution of the matrix exponential, a Mini-Max Polynomial Approximation (MMPA) method. This method is numerically stable for burn-up matrices with extremely short half-lived nuclides as the Chebyshev Rational Approximation Method (CRAM), and it has several advantages over CRAM. We also propose a multi-step calculation, a computational time reduction scheme of the MMPA method, which can perform simultaneously burn-up calculations with several time periods. The applicability of these methods has been theoretically and numerically proved for general burn-up matrices. The numerical verification has been performed, and it has been shown that these methods have high precision equivalent to CRAM

  8. Development of a set of benchmark problems to verify numerical methods for solving burnup equations

    International Nuclear Information System (INIS)

    Lago, Daniel; Rahnema, Farzad

    2017-01-01

    Highlights: • Description transmutation chain benchmark problems. • Problems for validating numerical methods for solving burnup equations. • Analytical solutions for the burnup equations. • Numerical solutions for the burnup equations. - Abstract: A comprehensive set of transmutation chain benchmark problems for numerically validating methods for solving burnup equations was created. These benchmark problems were designed to challenge both traditional and modern numerical methods used to solve the complex set of ordinary differential equations used for tracking the change in nuclide concentrations over time due to nuclear phenomena. Given the development of most burnup solvers is done for the purpose of coupling with an established transport solution method, these problems provide a useful resource in testing and validating the burnup equation solver before coupling for use in a lattice or core depletion code. All the relevant parameters for each benchmark problem are described. Results are also provided in the form of reference solutions generated by the Mathematica tool, as well as additional numerical results from MATLAB.

  9. PIE and separate effect test of high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, S.K.; Kim, D.H.

    2005-01-01

    To investigate the performance of a high burnup UO 2 fuel, the highest burnup fuel assembly in KOREA was transported to the PIE facility in KAERI. It was a 17·17 fuel assembly irradiated at the Ulchin Unit 2 PWR. The peak fuel rod average burnup was about 57MWd/kgU and locally 65MWd/kgU. The general PIE was performed to investigate the fuel rod irradiation performance. Fission gas release, burnup, oxide thickness, hydrogen pickup, CRUD, and density change were measured by destructive of non-destructive test. Microstructure change, bubble and pore size distributions were observed by optical microscopy, SEM and EPMA. All generated and available PIE results were used to verify high burnup fuel performance code INFRA. Several rods were cut for additional separate effect test. For the high burnup fission gas release behaviour analysis, annealing apparatus were developed and installed in hot cell and preliminary test was performed. In addition to current apparatus new induction furnace will be installed in hot cell to investigate the high temperature and transient fission gas release behaviour. Ring tensile test was performed to analyze the material property degradation which caused by the oxidation and hydride, and additional mechanical tests will be performed. (Author)

  10. Numerical solution of stiff burnup equation with short half lived nuclides by the Krylov subspace method

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Tatsumi, Masahiro; Sugimura, Naoki

    2007-01-01

    The Krylov subspace method is applied to solve nuclide burnup equations used for lattice physics calculations. The Krylov method is an efficient approach for solving ordinary differential equations with stiff nature such as the nuclide burnup with short lived nuclides. Some mathematical fundamentals of the Krylov subspace method and its application to burnup equations are discussed. Verification calculations are carried out in a PWR pin-cell geometry with UO 2 fuel. A detailed burnup chain that includes 193 fission products and 28 heavy nuclides is used in the verification calculations. Shortest half life found in the present burnup chain is approximately 30 s ( 106 Rh). Therefore, conventional methods (e.g., the Taylor series expansion with scaling and squaring) tend to require longer computation time due to numerical stiffness. Comparison with other numerical methods (e.g., the 4-th order Runge-Kutta-Gill) reveals that the Krylov subspace method can provide accurate solution for a detailed burnup chain used in the present study with short computation time. (author)

  11. MTR fuel element burn-up measurements by the reactivity method

    International Nuclear Information System (INIS)

    Zuniga, A.; Cuya, T.R.; Ravnik, M.

    2003-01-01

    Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

  12. Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States); Andrews, Nathan C., E-mail: nandrews@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States)

    2013-05-15

    Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

  13. Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

    International Nuclear Information System (INIS)

    Karahan, Aydın; Andrews, Nathan C.

    2013-01-01

    Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

  14. Substep methods for burnup calculations with Bateman solutions

    International Nuclear Information System (INIS)

    Isotalo, A.E.; Aarnio, P.A.

    2011-01-01

    Highlights: → Bateman solution based depletion requires constant microscopic reaction rates. → Traditionally constant approximation is used for each depletion step. → Here depletion steps are divided to substeps which are solved sequentially. → This allows piecewise constant, rather than constant, approximation for each step. → Discretization errors are almost completely removed with only minor slowdown. - Abstract: When material changes in burnup calculations are solved by evaluating an explicit solution to the Bateman equations with constant microscopic reaction rates, one has to first predict the development of the reaction rates during the step and then further approximate these predictions with their averages in the depletion calculation. Representing the continuously changing reaction rates with their averages results in some error regardless of how accurately their development was predicted. Since neutronics solutions tend to be computationally expensive, steps in typical calculations are long and the resulting discretization errors significant. In this paper we present a simple solution to reducing these errors: the depletion steps are divided to substeps that are solved sequentially, allowing finer discretization of the reaction rates without additional neutronics solutions. This greatly reduces the discretization errors and, at least when combined with Monte Carlo neutronics, causes only minor slowdown as neutronics dominates the total running time.

  15. Fuel removing method for high burnup fuel and device therefor

    International Nuclear Information System (INIS)

    Terakado, Shogo; Owada, Isao; Kanno, Yoshio; Aizawa, Sakue; Yamahara, Takeshi.

    1993-01-01

    A through hole is perforated at the center of a fuel rod in a cladding tube by a diamond drill in a water vessel. Further, the through hole is enlarged by the diamond drill. A pellet removing tool is attached to a drill chuck instead of the diamond drill. Then, the thin cylindrical fuel pellet remaining on the inner surface of the cladding tube is removed by using a pellet removing tool while applying vibrations. Subsequently, a wire brush having a slightly larger diameter than that of the inner diameter of the cladding tube is attached to the drill chuck and rotated to finish the inner surface, so that a small amount of pellets remained on the inner surface of the cladding tube is removed. Pellet powders in the water vessel are collected and recovered to the water container. This can remove high burnup fuels which are firmly sticked to the cladding tube, without giving thermal or mechanical influences on the cladding tube. (I.N.)

  16. Nondestructive, fast methods for burn-up study

    International Nuclear Information System (INIS)

    Schaechter, L.; Hacman, D.; Mot, O.

    1977-01-01

    Nondestructive methods, based on high resolution-spectrometry successfully applied at Institute for Atomic Physics are presented. These methods are preferred to destructive chemical methods; the latter being costly and lengthy and not suitable for statistical prediction of nuclear fuel behaviour. The following methods are developed: methods for determining the burn up of fuel elements and fuel assemblies; a method for determining the U 235 and Pu 239 contributions to the burn up and a code written in FORTRAN IV for numerical calculation of Pu 239 fission vs. burn up; a high precision method for burnup determination by adding burnable poison; a method for prediction of specific power distribution in the fuel elements of a research or power reactors; a method for determining the power output of the fuel element in an operating power reactor; a method for determining the content of Pu 239 of the fuel element irradiated in a reactor. The results which were obtained by these methods improved the fuel management at the VVR-S reactor at Institute for Atomic Physics, Bucharest and may be applied to other reactor types [fr

  17. Test of calorimetry for high burn-up plutonium

    International Nuclear Information System (INIS)

    Beets, C.; Carchon, R.; Fettweis, P.

    1984-01-01

    In recent times, the interest of applying calorimetry for safeguards purpose is steadily increasing. Calorimetric measurements have been performed on a set of high burn-up (25000 MWd/t) Pu samples, ranging in mass between 60 g and 2.5 kg Pu, distributed as PuO 2 powder embedded in stainless steel containers. The powers produced by these containers ranged between 0.8 W and 36 W. The calorimeter used was the Mound 150 type, and the isotopics and the Am content have been determined earlier by mass spectroscopy, completed with α and γ counting, and were later verified by the same methods. Watts/gram measurements were made on twelve 60 g samples of the same plutonium lot to demonstrate the Pu elemental and isotopic homogeneity, and hence, its suitability for subsequent NDA experiments. These samples were also measured in a stacked way to fill up the mass and wattage gaps between 60 g (0.8W) and 1 kg (14W). Calorimetric assay values, obtained with both isotopic measurements are discussed

  18. Electron angular distribution axial channeling

    International Nuclear Information System (INIS)

    Khokonov, A.Kh.; Khokonov, M.Kh.

    1989-01-01

    Angular distributions of ultra-relativistic electrons are calculated in the assumption about presence of statistical equilibrium. Analysis is based on numerical solution of Fokker-Planck type kinetic equation. It is shown that in contrast to case of amorphous medium, the multiple scattering at axial channeling of negative particles results in self-focusing of the initial beam particles and due to it number of electrons moving at an angles to the chain, which are smaller, than critical angle of channeling, may increase by several times as compared to the initial one

  19. Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

    Energy Technology Data Exchange (ETDEWEB)

    Catana, A.; Prodea, L. [RAAN, Institute for Nuclear Research, Arges (Romania); Danila, N.; Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica(Romania)

    2007-07-01

    The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up.

  20. Subchannel flow analysis in Candu and ACR pressure tubes with radial and axial diameter variation

    International Nuclear Information System (INIS)

    Catana, A.; Prodea, L.; Danila, N.; Prisecaru, I.; Dupleac, D.

    2007-01-01

    The Candu (Canada Deuterium Uranium) and ACR (Advanced Candu Reactor) are pressure tubes (PT) heavy water moderated reactors. Candu are heavy water and ACR are light water cooled reactors. The pressure tube is filled with 12 bundles, each consisting of 37 respectively 43 fuel rods. One Candu reactor is in operation at Cernavoda, Romania since 1996. ACR is a proposed advanced Candu. PT diameter variation has a significant impact on the thermal-hydraulic parameters. Almost all thermal-hydraulic parameters change, but some of them have a greater significance. In this work we have considered a set of radial and axial PT diameter variations both for Candu-600 and ACR-700 reactors using various types of fuel bundles. We can conclude the following: 1) some thermal-hydraulic parameters are significantly influenced: critical heat flux (CHF), pressure drop, or void fraction; 2) the most significant parameter CHF is worsening which reduces the safety margin; 3) some fuel types present a better thermal-hydraulic behavior; and 4) fuel bundles with fresh fuel or low burnup have a worse thermal-hydraulic behaviour than those at average burn-up

  1. IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2003-01-01

    grain boundary porosity in the central regions of the pellet. The back scattered electron image of the pellet periphery showed an area of high porosity as would be associated with re-structuring caused by formation of the Hobs. The depth of this region was typically 200 microns with traces of Hobs at greater depths. After transfer to Kjeller hot laboratories in Norway, three segments were cut from rod D5 33-25065. All were from regions of uniform burn-up as measured by axial gamma scanning. The sections were identified as D5:1, rod 7, D5:2, rod 8 and D5:3, rod 9. Whilst drilling rod 7 to insert a thermocouple, the drill broke and thermocouple insertion was abandoned. The rod was finally sealed with a clad elongation detector (EC). Rods 8 and 9 were successfully drilled and thermocouples inserted. They were then sealed with pressure transducers (PF). Rods 8 and 9 were loaded into positions 2 and 5 in IFA-597.2 (second loading) and irradiated in Halden for some 20 days in July 1995. After a small number of power ramps, rod 9 failed and the assembly withdrawn. During this time, useful data were generated on centreline temperature as a function of power. Rod 9 was removed and replaced by rod 7. The assembly was returned to the reactor as IFA-597.3 (third loading); the irradiation started in January 1997 and continued to May of that year having accrued a further ∼2 MWd/kg UO 2 . Data obtained included, centreline temperature as a function of power and burn-up, (rod 8), FGR from the increase in rod internal pressure (rod 8) and clad elongation (rod 7). The assembly was discharged and transported to Kjeller for PIE. FGR of 12.6% and 15.8% were measured from puncturing and gas extraction from rods 7 and 8 respectively. Further PIE was carried out on rod 8 after it had been transported to Studsvik. Here, additional ceramography was performed, axial gamma scans, radial scans of selected fission products and the radial porosity distribution. On account of the high burn-up of

  2. Axially alignable nuclear fuel pellets

    International Nuclear Information System (INIS)

    Johansson, E.B.; Klahn, D.H.; Marlowe, M.O.

    1978-01-01

    An axially alignable nuclear fuel pellet of the type stacked in end-to-end relationship within a tubular cladding is described. Fuel cladding failures can occur at pellet interface locations due to mechanical interaction between misaligned fuel pellets and the cladding. Mechanical interaction between the cladding and the fuel pellets loads the cladding and causes increased cladding stresses. Nuclear fuel pellets are provided with an end structure that increases plastic deformation of the pellets at the interface between pellets so that lower alignment forces are required to straighten axially misaligned pellets. Plastic deformation of the pellet ends results in less interactions beween the cladding and the fuel pellets and significantly lowers cladding stresses. The geometry of pellets constructed according to the invention also reduces alignment forces required to straighten fuel pellets that are tilted within the cladding. Plastic deformation of the pellets at the pellet interfaces is increased by providing pellets with at least one end face having a centrally-disposed raised area of convex shape so that the mean temperature and shear stress of the contact area is higher than that of prior art pellets

  3. Exact axially symmetric galactic dynamos

    Science.gov (United States)

    Henriksen, R. N.; Woodfinden, A.; Irwin, J. A.

    2018-05-01

    We give a selection of exact dynamos in axial symmetry on a galactic scale. These include some steady examples, at least one of which is wholly analytic in terms of simple functions and has been discussed elsewhere. Most solutions are found in terms of special functions, such as associated Lagrange or hypergeometric functions. They may be considered exact in the sense that they are known to any desired accuracy in principle. The new aspect developed here is to present scale-invariant solutions with zero resistivity that are self-similar in time. The time dependence is either a power law or an exponential factor, but since the geometry of the solution is self-similar in time we do not need to fix a time to study it. Several examples are discussed. Our results demonstrate (without the need to invoke any other mechanisms) X-shaped magnetic fields and (axially symmetric) magnetic spiral arms (both of which are well observed and documented) and predict reversing rotation measures in galaxy haloes (now observed in the CHANG-ES sample) as well as the fact that planar magnetic spirals are lifted into the galactic halo.

  4. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F; Werle, H [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

  5. Nuclear criticality safety: general. 5. Reactivity Effect of Burnable Absorbers in Burnup Credit for the CASTOR X/32S Storage and Transport Cask

    International Nuclear Information System (INIS)

    Rombough, Charles T.; Lancaster, Dale B.; Diersch, Rudolf; Spilker, Harry

    2001-01-01

    When considering burnup credit in the licensing of storage and transportation casks, a significant effect is whether or not the burned fuel was depleted with burnable absorbers present. This paper presents the results of detailed calculations to quantitatively determine the burnable absorber effect for the CASTOR X/32S transport cask, which assumes burnup of the fuel in the criticality analysis. A radial view of the CASTOR X/32S cask is shown in Fig. 1. This is the actual plot of the geometry as modeled in KENO V.a. Note that there are no water-filled flux traps and the assemblies are tightly packed. This reduces the overall dimensions of the cask for a given number of fuel assemblies. Reactivity is held down by borated aluminum plates between the fuel assemblies and by placing absorber rod modules (ARMs) in the guide tubes of selected assemblies. If burnup of the fuel is not considered and the initial enrichment is 5.0 wt% 235 U, then 28 of the 32 fuel assemblies must contain an ARM to maintain a k eff 3 ; 4. moderator temperature of 604 K; 5. cooling time of 9.5 yr; 6. specific power of 60 W/g of U metal; 7. conservative axial and radial burnup shape distribution; 8. Westinghouse BP material containing 12.5 wt% B 4 C. Using the model described earlier, calculations were performed with varying numbers of BP fingers inserted for different exposure times. The results are shown in Tables I and II. The 1 s statistical error in these results is σ equals ±0.05%. Note that the BP finger and exposure effects decrease with fuel burnup and the effect is smaller when the cask contains ARMs. Conservatively combining the results from Tables I and II and interpolating, we can equate fewer BP fingers with longer BP exposure time as shown in Table III. The Table III results were checked by running the actual cases (for example, 20 BP fingers for 24 GWd/tonne exposure) to verify that the k eff 's for the cask were always less than the base-case values. These results can also be

  6. Development of an extended-burnup Mark B design. Second semiannual progress report, January-June 1979

    International Nuclear Information System (INIS)

    1979-11-01

    The immediate goal of the DOE/AP and L/B and W project is to extend the burnup of light water reactor fuel assemblies beyond present limits to 50,000 MWd/mtU batch average burnup. Fuel management plans and fuel designs are being directed to attain the increased burnup limits. Lead-test assemblies of extended-burnup designs will be manufactured, irradiated in a commercial pressurized water reactor, and examined to support extended-burnup fuel cycles. This report, covering the period from January through June 1979, is the second semiannual progress report for the program. Efforts have included analyses of extended-burnup fuel cycles, developed of both annular fuel pellet and segmented rod designs, and design of a nondestructive post-irradiation examination system

  7. DELIGHT-B/REDEL, point reactivity burnup code for high-temperature gas-cooled reactor cells

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Watanabe, Takashi.

    1977-03-01

    Code DELIGHT-2 was previously developed to analyze cell burnup characteristics and to produce few-group constants for core burnup calculation in high-temperature gas-cooled reactors. In the code, burnup dependency of the burnable poison, boron-10, is considered with the homogeneous model of space. In actuality, however, the burnable poison is used as homogeneous rods or uniform rods of small granular poison and graphite, to control the reactivity and power distribution. Precise analysis of the burnup characteristics is thus difficult because of the heterogeneity due to the configuration of poison rods. In cell burnup calculation, the DELIGHT-B, which is a modification of DELIGHT-2, takes into consideration this heterogeneous effect. The auxiliary code REDEL, a reduction of DELIGHT-B, used in combination with 3 dimensional diffusion code CITATION, is for core burnup calculation with the macro-scopic cross section model. (auth.)

  8. Mesoscopic approach to describe high burn-up fuel behaviour

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1999-01-01

    The grain sub-division and the rim structure formation are new phenomena for LWR fuel engineering. The consequence of these are now under investigation in several international programs such as HBRP (High Burnup Rim Project) of CRIEPI, NFIR of EPRI, and EdF/CEA program in France. The theoretical understanding of this phenomenon is underway. Here, the process is peculiar in the following points; (1) majority of the domain of the material are changed to a new morphology after the restructuring, (2) the final size of the new grains is around 0.1 μm which is neither atomic scale nor macroscopic scale. (3) the morphology of the restructured domain indicates fractal like feature which indicates complex process is under-taken. From the first feature, the process is similar to phase transitions or metallographic transformations. However, as the crystallographic structure has no change before and after the restructuring, it is not the phase transition nor the transformation of atomic scale instability. The focus could be put on the material transport of mesoscopic scale which create the peculiar morphology. Indeed there are flows of energy and disturbances in crystallographic structure in nuclear materials on duty. Although the fission energy is 10 4 larger than the formation energy of the defects, thanks to the stability of the selected material, most of energy is thermalized without crystallographic instability. Little remained energy creates flows of disturbances and the new structure is a consequence of ordering process driven by these flows of disturbances. Therefore this phenomenon is a good example to study cooperative ordering process in physics of materials. This paper presents some of present understandings of the rim structure formation based on the mesoscopic mechanistic theories. Possible future development is also proposed (author) (ml)

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

    International Nuclear Information System (INIS)

    Itagaki, N.; Ohira, K.; Tsuda, K.; Fischer, G.; Ota, T.

    1998-01-01

    UO 2 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 UO 2 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)

  10. Effect of core burnup on the dynamic behavior of fast reactors

    International Nuclear Information System (INIS)

    Ilberg, D.; Saphier, D.; Yiftah, S.

    1977-01-01

    Performance of a dynamic analysis, taking burnup changes into account, requires fission-product nuclear data of relatively small uncertainty, suitable burnup calculation models, and dynamic computer programs. These were prepared and used with the following results: (1) Significant changes in static and dynamic parameters were observed when investigating the effect of burnup. These changes were found to be larger than differences introduced by the uncertainty of the fission-product nuclear data. (2) A one-dimensional burnup computer program was prepared. It was found that a burnup model based on the generalized radioactive decay scheme is suitable for accurate fast reactor calculations. (3) Space-time dynamic calculations of fast reactors having different burnup levels were performed. The stability difference between ''clean'' and high burnup cores is greater when local rather than uniform perturbations are inserted along the entire core length. The magnitude by which the ''end-of-life'' core increases the transient excursion over that of the clean core depends on the particular region in which the perturbation is inserted. The end-of-life core will magnify the transient excursion more than the clean core whenever the perturbation is inserted into a region having a higher adjoint flux level than that of the clean core. However, when a reactor safety system operates successfully, the difference in the temperature transient of the clean and end-of-life cores will be relatively small. It is suggested that only the analysis of large local perturbations be performed for end-of-life cores as well as for clean cores in the safety evaluation of fast reactors

  11. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

    2016-10-21

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

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

    International Nuclear Information System (INIS)

    Lebrun, A.; Bignan, G.

    2001-01-01

    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

  13. Axial vector mass spectrum and mixing angles

    International Nuclear Information System (INIS)

    Caffarelli, R.V.; Kang, K.

    1976-01-01

    Spectral sum rules of the axial-vector current and axial-vector current-pseudoscalar field are used to study the axial-vector mass spectrum and mixing angles, as well as the decay constants and mixing angles of the pseudoscalar mesons. In general, the result is quite persuasive for the existence of the Jsup(PC) = 1 ++ multiplet in which one has a canonical D-E mixing. (Auth.)

  14. Electric machines with axial magnetic flux

    Science.gov (United States)

    Nuca, I.; Ambros, T.; Burduniuc, M.; Deaconu, S. I.; Turcanu, A.

    2018-01-01

    The paper contains information on the performance of axial machines compared to cylindrical ones. At the same time, various constructive schemes of synchronous electromechanical converters with permanent magnets and asynchronous with short-circuited rotor are presented. In the developed constructions, the aim is to maximize the usage of the material of the stator windings. The design elements of the axial machine magnetic system are presented. The FEMM application depicted the array of the magnetic field of an axial machine.

  15. Centrifugal and axial compressor control

    CERN Document Server

    McMillan, Gregory K

    2009-01-01

    Control engineers, mechanical engineers and mechanical technicians will learn how to select the proper control systems for axial and centrifugal compressors for proper throughput and surge control, with a particular emphasis on surge control. Readers will learn to understand the importance of transmitter speed, digital controller sample time, and control valve stroking time in helping to prevent surge. Engineers and technicians will find this book to be a highly valuable guide on compressor control schemes and the importance of mitigating costly and sometimes catastrophic surge problems. It can be used as a self-tutorial guide or in the classroom with the book's helpful end-of-chapter questions and exercises and sections for keeping notes.

  16. Axial channeling in electron diffraction

    International Nuclear Information System (INIS)

    Ichimiya, A.; Lehmpfuhl, G.

    1978-01-01

    Kossel patterns from Silicon and Niobium were obtained with a convergent electron beam. An intensity maximum in the direction of the zone axes [001] and [111] of Nb was interpreted as axial channeling. The intensity distribution in Kossel patterns was calculated by means of the Bloch wave picture of the dynamical theory of electron diffraction. Particularly zone axis patterns were calculated for different substance-energy combinations and they were compared with experimental observations. The intensity distribution in the calculated Kossel patterns was very sensitive to the model of absorption and it was found that a treatment of the absorption close to the model of Humphreys and Hirsch [Phil. Mag. 18, 115 (1968)] gave the best agreement with the experimental observations. Furthermore it is shown which Bloch waves are important for the intensity distribution in the Kossel patterns, how they are absorbed and how they change with energy. (orig.) [de

  17. Axial channeling of uttrarelativistic electrons

    Energy Technology Data Exchange (ETDEWEB)

    Telegin, V.I.; Khokonov, M.Kh. (Moskovskij Gosudarstvennyj Univ. (USSR). Nauchno-Issledovatel' skij Inst. Yadernoj Fiziki)

    1982-07-01

    The dynamics of motion of ultrarelativistic electrons under axial channeling conditions is investigated. The analysis is based on the solution of the kinetic equation obtained recently by Beloshitsky and Kumakhov. The particle dechanneling function is investigated as depending on the type of a crystal, particle energy and angle of entrance into the single crystal. It is found that for most of the beam the major diffusion mechanism is scattering by electrons. It is shown that an optimal depth range exists for which the fraction of channeled particles sharply increases at the expense of the quasi-channeled particles. In a number of cases the dechanneling length for crystals with high atomic numbers may be greater than that of light elements.

  18. Axial channeling of uttrarelativistic electrons

    International Nuclear Information System (INIS)

    Telegin, V.I.; Khokonov, M.Kh.

    1982-01-01

    The dynamics of motion of ultrarelativistic electrons under axial channeling conditions is investigated. The analysis is based on the solution of the kinetic equation obtained recently by Beloshitsky and Kumakhov. The particle dechanneling function is investigated as depending on the type of a crystal, particle energy and angle of entrance into the single crystal. It is found that for most of the beam the major diffusion mechanism is scattering by electrons. It is shown that an optimal depth range exists for which the fraction of channeled particles sharply increases at the expense of the quasi-channeled particles. In a number of cases the dechanneling length for crystals with high atomic numbers may be greater than that of light elements

  19. Bone Disease in Axial Spondyloarthritis.

    Science.gov (United States)

    Van Mechelen, Margot; Gulino, Giulia Rossana; de Vlam, Kurt; Lories, Rik

    2018-05-01

    Axial spondyloarthritis is a chronic inflammatory skeletal disorder with an important burden of disease, affecting the spine and sacroiliac joints and typically presenting in young adults. Ankylosing spondylitis, diagnosed by the presence of structural changes to the skeleton, is the prototype of this disease group. Bone disease in axial spondyloarthritis is a complex phenomenon with the coexistence of bone loss and new bone formation, both contributing to the morbidity of the disease, in addition to pain caused by inflammation. The skeletal structural changes respectively lead to increased fracture risk and to permanent disability caused by ankylosis of the sacroiliac joints and the spine. The mechanism of this new bone formation leading to ankylosis is insufficiently known. The process appears to originate from entheses, specialized structures that provide a transition zone in which tendon and ligaments insert into the underlying bone. Growth factor signaling pathways such as bone morphogenetic proteins, Wnts, and Hedgehogs have been identified as molecular drivers of new bone formation, but the relationship between inflammation and activation of these pathways remains debated. Long-standing control of inflammation appears necessary to avoid ankylosis. Recent evidence and concepts suggest an important role for biomechanical factors in both the onset and progression of the disease. With regard to new bone formation, these processes can be understood as ectopic repair responses secondary to inflammation-induced bone loss and instability. In this review, we discuss the clinical implications of the skeletal changes as well as the underlying molecular mechanisms, the relation between inflammation and new bone formation, and the potential role of biomechanical stress.

  20. Origin of axial current in scyllac

    International Nuclear Information System (INIS)

    Sugisaki, K.

    1975-12-01

    The origin of the axial current observed in Scyllac (a high beta stellarator experiment) is discussed. A shaped coil and/or helical winding produce rotational transform which links magnetic lines of force to the plasma column and the axial current is induced electromagnetically. This phenomenon is inherent in a pulsed high-beta stellarator. The rotational transform produced by the induced axial current is much smaller than that associated with the l = 1, 0 equilibrium fields. The effect of the axial current on the equilibrium and stability of the plasma column is thus small. It is also shown that the magnetic field shear near a plasma surface is very strong

  1. Radial and axial compression of pure electron

    International Nuclear Information System (INIS)

    Park, Y.; Soga, Y.; Mihara, Y.; Takeda, M.; Kamada, K.

    2013-01-01

    Experimental studies are carried out on compression of the density distribution of a pure electron plasma confined in a Malmberg-Penning Trap in Kanazawa University. More than six times increase of the on-axis density is observed under application of an external rotating electric field that couples to low-order Trivelpiece-Gould modes. Axial compression of the density distribution with the axial length of a factor of two is achieved by controlling the confining potential at both ends of the plasma. Substantial increase of the axial kinetic energy is observed during the axial compression. (author)

  2. CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback

    International Nuclear Information System (INIS)

    Ahnert, Carol; Aragones, Jose M.

    1983-01-01

    1 - Description of problem or function: CARMEN is a system of programs developed for the neutronic calculation of PWR cycles. It includes the whole chain of analysis from cell calculations to core calculations with burnup. The core calculations are based on diffusion theory with cross sections depending on the relevant space-dependent feedback effects which are present at each moment along the cycles. The diffusion calculations are in one, two or three dimensions and in two energy groups. The feedback effects which are treated locally are: burnup, water density, power density and fission products. In order to study in detail these parameters the core should be divided into as many zones as different cross section sets are expected to be required in order to reproduce reality correctly. A relevant difference in any feedback parameter between zones produces different cross section sets for the corresponding zones. CARMEN is also capable to perform the following calculations: - Multiplication factor by burnup step with fixed boron concentration - Buckling and control rod insertion - Buckling search by burnup step - Boron search by burnup step - Control rod insertion search by burnup step. 2 - Method of solution: The cell code (LEOPARD-TRACA) generates the fuel assembly cross sections versus burnup. This is the basic library to be used in the CARMEN code proper. With a planar distribution guess for power density, water density and fluxes, the macroscopic cross sections by zone are calculated by CARMEN, and then a diffusion calculation is done in the whole geometry. With the distribution of power density, heat accumulated in the coolant and the thermal and fast fluxes determined in the diffusion calculation, CARMEN calculates the values of the most relevant parameters that influence the macroscopic cross sections by zone: burnup, water density, effective fuel temperature and fission product concentrations. If these parameters by zone are different from the reference

  3. Effect of core configuration on the burnup calculations of MTR research reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Amin, E.H.; Sakr, A.M.

    2014-01-01

    Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

  4. Estimation of the impact of manufacturing tolerances on burn-up calculations using Monte Carlo techniques

    Energy Technology Data Exchange (ETDEWEB)

    Bock, M.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, Garching (Germany). Forschungszentrum

    2012-11-01

    In recent years, the availability of computing resources has increased enormously. There are two ways to take advantage of this increase in analyses in the field of the nuclear fuel cycle, such as burn-up calculations or criticality safety calculations. The first possible way is to improve the accuracy of the models that are analyzed. For burn-up calculations this means, that the goal to model and to calculate the burn-up of a full reactor core is getting more and more into reach. The second way to utilize the resources is to run state-of-the-art programs with simplified models several times, but with varied input parameters. This second way opens the applicability of the assessment of uncertainties and sensitivities based on the Monte Carlo method for fields of research that rely heavily on either high CPU usage or high memory consumption. In the context of the nuclear fuel cycle, applications that belong to these types of demanding analyses are again burn-up and criticality safety calculations. The assessment of uncertainties in burn-up analyses can complement traditional analysis techniques such as best estimate or bounding case analyses and can support the safety analysis in future design decisions, e.g. by analyzing the uncertainty of the decay heat power of the nuclear inventory stored in the spent fuel pool of a nuclear power plant. This contribution concentrates on the uncertainty analysis in burn-up calculations of PWR fuel assemblies. The uncertainties in the results arise from the variation of the input parameters. In this case, the focus is on the one hand on the variation of manufacturing tolerances that are present in the different production stages of the fuel assemblies. On the other hand, uncertainties that describe the conditions during the reactor operation are taken into account. They also affect the results of burn-up calculations. In order to perform uncertainty analyses in burn-up calculations, GRS has improved the capabilities of its general

  5. Modelling of phenomena associated with high burnup fuel behaviour during overpower transients

    International Nuclear Information System (INIS)

    Sills, H.E.; Langman, V.J.; Iglesias, F.C.

    1995-01-01

    Phenomena of importance to the behaviour of high burnup fuel subjected to conditions of rapid overpower (i.e., LWR RIAs) include the change in cladding material properties due to irradiation, pellet-clad interaction (PCI) and 'rim' effects associated with the periphery of high burnup fuel. 'Rim' effects are postulated to be caused by changes in fuel morphology at high burnup. Typical discharge burnups for CANDU fuel are low compared to LWRs. Maximum linear ratings for CANDU fuel are higher than those for LWRs. However, under normal operating conditions, the Zircaloy-4 clad of the CANDU fuel is collapsed onto the fuel stack. Thus, the CANDU fuel performance codes model the transient behaviour of the fuel-to-clad interface and are capable of assessing the potential for pellet-clad mechanical interaction (PCMI) failures for a wide range of overpower conditions. This report provides a discussion of the modelling of the phenomena of importance to high burnup fuel behaviour during rapid overpower transients. (author)

  6. Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1991-04-01

    To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs

  7. A simplified burnup calculation strategy with refueling in static molten salt reactor

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Gupta, Anurag; Krishnani, P.D.

    2015-01-01

    Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233 Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)

  8. MCNPX and MCB coupled methodology for the burnup calculation of the KIPT accelerator driven subcritical system

    International Nuclear Information System (INIS)

    Zhong, Z.; Gohar, Y.; Talamo, A.

    2009-01-01

    Argonne National Laboratory (ANL) of USA and Kharkov Inst. of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical facility (ADS). The facility will be utilized for basic research, medical isotopes production, and training young nuclear specialists. The burnup methodology and analysis of the KIPT ADS are presented in this paper. MCNPX and MCB Monte Carlo computer codes have been utilized. MCNPX has the capability of performing electron, photon and neutron coupled transport problems, but it lacks the burnup capability for driven subcritical systems. MCB has the capability for performing the burnup calculation of driven subcritical systems, while it cannot transport electrons. A calculational methodology coupling MCNPX and MCB has been developed, which can exploit the electrons transport capability of MCNPX for neutron production and the burnup capability of MCB for driven subcritical systems. In this procedure, a neutron source file is generated using MCNPX transport calculation, preserving the neutrons yield from photonuclear reactions initiated by electrons, and this source file is utilized by MCB for the burnup analyses with the same geometrical model. In this way, the ADS depletion calculation can be accurately. (authors)

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

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1995-01-01

    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

  10. Past experience and future needs for the use of burnup credit in LWR fuel storage

    International Nuclear Information System (INIS)

    Boyd, W.A.; Wrights, G.N.

    1987-01-01

    To achieve improved fuel economics and reduce the amount of fuel discharged annually, utilities are engaging in fuel management strategies that will achieve higher discharge burnups for their fuel assemblies. Although burnup credit methodologies have been developed and spent-fuel racks have been licensed, burnup credit fuel storage racks are not the answer for all utilities. Off-site and out-of-pool spent-fuel storage may be more appropriate. This is leading to the development of dry spent-fuel storage and shipping casks. Cask designs with spent-fuel storage capability between 20 and 32 assemblies are being developed by several vendors. The US Dept. of Energy is also funding work by VEPCO. Westinghouse is currently licensing its dry storage cask, developing a shipping cask for the domestic market, and is involved in a joint venture to develop a cask for the international market. Although methods of taking credit for fuel burnup in spent-fuel storage racks have been developed and licensed, use of these methods on dry spent-fuel storage and shipping casks can lead to new issues. These issues arise because the excess reactivity margin that is inherent in a burnup credit spent-fuel storage rack criticality analysis will not be available in a dry cask analysis

  11. Design and analytic evaluation of a rim effect reduction type LWR fuel for extending burnup

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo; Kameyama, Takanori; Kinoshita, Motoyasu

    1991-01-01

    We have designed a new concept fuel design 'Rim effect reduction type fuel' which has thin natural UO 2 layer on surface of a UO2 pellet. Our neutronic analyses with ANRB code show this fuel design can reduce rim effect (burnup at plelet rim) by about 30 GWd/t comparing a normal fuel. It is known that a high burnup fuel has different microstructure from as-fabricated one at fuel rim (which is called as rim region) due to rim effect. Therefore this fuel design can expect smaller rim region than a normal fuel. Our fuel performance analyses with EIMUS code show this fuel design can reduce fuel center temperature at high burnup if thermal conductivity of fuel pellet decreases with burnup in inverse proportion. However, this fuel design increases fuel center temperature at low and middle burnup than a normal fuel due to increase of thermal power density at pellet center. Additionally Irradiation experiment of this fuel design can be considered to offer important data which make clear the relation between rim effect and fuel performance. (author)

  12. Review of the effects of burnup on the thermal conductivity of UO2

    International Nuclear Information System (INIS)

    Lokken, R.O.; Courtright, E.L.

    1976-01-01

    The general trends which relate changes in thermal conductivity of UO 2 fuel as a function of temperature and burnup can be summarized as follows: (1) At temperatures below 500 0 C, reductions in UO 2 thermal conductivity relative to the unirradiated values can be expected up to a saturation level of approximately 10 19 fissions/cc. (2) At temperatures above 500 0 C, the thermal conductivity will undergo little change at low burnups, (less than 10 19 fissions/cc) but at higher exposures some decrease can be expected which should, in turn, diminish with increasing temperature. (3) A review of the data reported by Berman on the ThO 2 --UO 2 fuel indicates that the basic behavior is the same as for UO 2 in the temperature range of major interest. The applicability of this data to LWR UO 2 fuel is somewhat questionable because of basic physical property differences, and limited data on irradiation effects, and would not seem to support concerns that the effects of burnup on thermal conductivity for LWR fuel may be of more significance than currently believed. (4) A mathematical expression of the type proposed by Daniel and Cohen seems to provide a reasonable approximation for the behavioral trends reported in the literature which relate changes in thermal conductivity to increasing burnup in certain temperature regimes. Calculations indicate that only small incremental increases in the fuel centerline temperature might be expected if burnup effects are taken into account

  13. Experimental and theoretical burnup investigations on model arrangements with solid burnable poisons

    International Nuclear Information System (INIS)

    Ahlf, J.; Anders, D.; Greim, L.; Knoth, J.; Kolb, M.; Mittelstaedt, B.; Mueller, A.; Schwenke, H.

    1975-01-01

    It is the scope of the two experiments here to improve the methods for computation and measurement as well as the experimental technique appropriate to predict the burnable poison rod burn-up with sufficient accuracy. In the first experiment two nine-rod bundles in a 3 x 3 arrangement are irradiated during several irradiation periods in the research reactor Geesthacht. Each bundle consists of eight outer rods containing fuel and one inner rod containing poison (B 10 or Cd 113). The burn-up of the fuel and the burnable poison is measured by non-destructive methods after each irradiation period and then compared with results of a burn-up calculation. In the second experiment two poison rods with different cadmium concentrations and one rod containing boron are irradiated during several irradiation periods in the research reactor Geesthacht. The burn-up is determined after each irradiation period by reactivity measurements and its result compared to computed effective absorption cross-sections of the rods by aid of a calibration curve. For both experiments the experimental and theoretical results for the poison burn-up are found to be within the error limits of the measurements. (orig.) [de

  14. Experimental and theoretical investigations on solid burnable poison burnup of model arrangements

    International Nuclear Information System (INIS)

    Ahlf, J.; Anders, D.; Greim, L.; Knoth, J.; Kolb, M.; Mittelstaedt, B.; Mueller, A.; Schwenke, H.

    1975-01-01

    It is the scope of the two experiments reported here to improve the methods for computation and measurement as well as the experimental technique appropriate to predict the burnable poison rod burn-up with sufficient accuracy. In the first experiment two nine-rod bundles in a 3 x 3 arrangement are irradiated during several irradiation periods in the research reactor Geesthacht. Each bundle consists of eight outer rods containing fuel and one inner rod containing poison (B 10 or Cd 113). The burn-up of the fuel and the burnable poison is measured by non-destructive methods after each irradiation period and then compared with results of a burn-up calculation. In the second experiment two poison rods with different cadmium concentrations and one rod containing boron are irradiated during several irradiation periods in the research reactor Geesthacht. The burn-up is determined after each irradiation period by reactivity measurements and its result compared to computed effective absorption cross-sections of the rods by aid of a calibration curve. For both experiments the experimental and theoretical results for the poison burn-up are found to be within the error limits of the measurements. (orig.) [de

  15. Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

    Energy Technology Data Exchange (ETDEWEB)

    Strain, R V; Johnson, C E

    1978-05-01

    Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

  16. High frequency acoustic microscopy for the determination of porosity and Young's modulus in high burnup uranium dioxide nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marchetti, M. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O. Box 2340 76125 Karlsruhe (Germany); University of Montpellier, IES, UMR 5214, F-34000, Montpellier (France); Laux, D. [University of Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France); Cappia, F. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O. Box 2340 76125 Karlsruhe (Germany); Technische Universitaet Muenchen, Department of Nuclear Engineering, Boltzmannstrasse 15, 85747 Garching bei Munchen (Germany); Laurie, M.; Van Uffelen, P.; Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O. Box 2340 76125 Karlsruhe (Germany); Despaux, G. [University of Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France)

    2015-07-01

    During irradiation UO{sub 2} nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of the porosity and of elastic properties in high burnup UO{sub 2} pellet can be investigated via high frequency acoustic microscopy. Ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A 67 MWd/kgU UO{sub 2} pellet was characterized using the acoustic microscope installed in the hot cells of the Institute of Transuranium Elements: 90 MHz frequency was applied, methanol was used as coupling liquid and VR was measured at different radial positions. By comparing the porosity values obtained via acoustic microscopy with those determined using ceramographic image analysis a good agreement was found, especially in the areas close to the centre. In addition Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile. (authors)

  17. High frequency acoustic microscopy for the determination of porosity and Young's modulus in high burnup uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Marchetti, M.; Laux, D.; Cappia, F.; Laurie, M.; Van Uffelen, P.; Rondinella, V.V.; Despaux, G.

    2015-01-01

    During irradiation UO 2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of the porosity and of elastic properties in high burnup UO 2 pellet can be investigated via high frequency acoustic microscopy. Ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A 67 MWd/kgU UO 2 pellet was characterized using the acoustic microscope installed in the hot cells of the Institute of Transuranium Elements: 90 MHz frequency was applied, methanol was used as coupling liquid and VR was measured at different radial positions. By comparing the porosity values obtained via acoustic microscopy with those determined using ceramographic image analysis a good agreement was found, especially in the areas close to the centre. In addition Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile. (authors)

  18. Stochastic quantization for the axial model

    International Nuclear Information System (INIS)

    Farina, C.; Montani, H.; Albuquerque, L.C.

    1991-01-01

    We use bosonization ideas to solve the axial model in the stochastic quantization framework. We obtain the fermion propagator of the theory decoupling directly the Langevin equation, instead of the Fokker-Planck equation. In the Appendix we calculate explicitly the anomalous divergence of the axial-vector current by using a regularization that does not break the Markovian character of the stochastic process

  19. Health and imaging outcomes in axial spondyloarthritis

    NARCIS (Netherlands)

    Machado, P.M.

    2016-01-01

    This thesis focuses on the assessment and monitoring of health and imaging outcomes in axial spondyloarthritis (SpA) and the relationship between these outcomes. Four major contributions to the understanding and management of axial SpA were made: 1) the improvement and facilitation of the assessment

  20. Some implications of batch average burnup calculations on predicted spent fuel compositions

    International Nuclear Information System (INIS)

    Alexander, C.W.; Croff, A.G.

    1984-01-01

    The accuracy of using batch-averaged burnups to determine spent fuel characteristics (such as isotopic composition, activity, etc.) was examined for a typical pressurized-water reactor (PWR) fuel discharge batch by comparing characteristics computed by (a) performing a single depletion calculation using the average burnup of the spent fuel and (b) performing separate depletion calculations based on the relative amounts of spent fuel in each of twelve burnup ranges and summing the results. The computations were done using ORIGEN 2. Procedure (b) showed a significant shift toward a greater quantity of the heavier transuranics, which derive from multiple neutron captures, and a corresponding decrease in the amounts of lower transuranics. Those characteristics which derive primarily from fission products, such as total radioactivity and total thermal power, are essentially identical for the two procedures. Those characteristics that derive primarily from the heavier transuranics, such as spontaneous fission neutrons, are underestimated by procedure (a)