WorldWideScience

Sample records for continuous energy burnup

  1. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  2. 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.

  3. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  4. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  5. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  6. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  7. Continuous-Energy Data Checks

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, Wim [Radioprotection and Nuclear Safety Institute, Fontenay-aux-Roses (France); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCartney, Austin Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald Kent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-25

    The purpose of this report is to provide an overview of all Quality Assurance tests that have to be performed on a nuclear data set to be transformed into an ACE formatted nuclear data file. The ACE file is capable of containing different types of data such as continuous energy neutron data, thermal scattering data, etc. Within this report, we will limit ourselves to continuous energy neutron data.

  8. Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

    Directory of Open Access Journals (Sweden)

    Kępisty Grzegorz

    2015-09-01

    Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

  9. Configurational space continuity and free energy calculations

    CERN Document Server

    Tian, Pu

    2016-01-01

    Free energy is arguably the most importance function(al) for understanding of molecular systems. A number of rigorous and approximate free energy calculation/estimation methods have been developed over many decades. One important issue, the continuity of an interested macrostate (or path) in configurational space, has not been well articulated, however. As a matter of fact, some important special cases have been intensively discussed. In this perspective, I discuss the relevance of configurational space continuity in development of more efficient and reliable next generation free energy methodologies.

  10. Continuous Energy Photon Transport Implementation in MCATK

    Energy Technology Data Exchange (ETDEWEB)

    Adams, Terry R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trahan, Travis John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sweezy, Jeremy Ed [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nolen, Steven Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hughes, Henry Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pritchett-Sheats, Lori A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Werner, Christopher John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-31

    The Monte Carlo Application ToolKit (MCATK) code development team has implemented Monte Carlo photon transport into the MCATK software suite. The current particle transport capabilities in MCATK, which process the tracking and collision physics, have been extended to enable tracking of photons using the same continuous energy approximation. We describe the four photoatomic processes implemented, which are coherent scattering, incoherent scattering, pair-production, and photoelectric absorption. The accompanying background, implementation, and verification of these processes will be presented.

  11. Calibration of burnup monitor installed in Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Naito, Hirofumi; Hirota, Masanari [Japan Nuclear Fuel Co. Ltd., Rokkasho, Aomori (Japan); Natsume, Koichiro [Isogo Engineering Center, Toshiba Corporation, Yokohama, Kanagawa (Japan); Kumanomido, Hironori [Nuclear Engineering Laboratory, Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2000-06-01

    Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid. (author)

  12. Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan

    2015-09-15

    Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

  14. A Simple Global View of Fuel Burnup

    Science.gov (United States)

    Sekimoto, Hiroshi

    2017-01-01

    Reactor physics and fuel burnup are discussed in order to obtain a simple global view of the effects of nuclear reactor characteristics to fuel cycle system performance. It may provide some idea of free thinking and overall vision, though it is still a small part of nuclear energy system. At the beginning of this lecture, governing equations for nuclear reactors are presented. Since the set of these equations is so big and complicated, it is simplified by imposing some extreme conditions and the nuclear equilibrium equation is derived. Some features of future nuclear equilibrium state are obtained by solving this equation. The contribution of a nucleus charged into reactor core to the system performance indexes such as criticality is worth for understanding the importance of each nuclide. It is called nuclide importance and can be evaluated by using the equations adjoint to the nuclear equilibrium equation. Examples of some importances and their application to criticalily search problem are presented.

  15. 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.

  16. 1D Burnup Calculation of Fusion-Fission Hybrid Energy Reactor%聚变-裂变混合能源堆一维计算模型燃耗分析

    Institute of Scientific and Technical Information of China (English)

    李茂生; 师学明; 伊炜伟

    2012-01-01

    Fusion-fission hybrid energy reactor is driven by Tokamak fusion source for energy production. Its subcritical zone uses the natural uranium as fuel and water as coolant. The neutron multiplication constant keff, energy multiplication factor M and tritium breeding ratio TBR of the ID hybrid energy reactor model were calculated by transport burnup code MCORGS. The neutron spectrum and nuclear density changing as a function of time show the characteristics of the hybrid energy reactors, which differs from the hybrid reactor for breed nuclear fuel and for spent fuel transmutation. The definition and results may be a reference to the other conceptual analysis.%聚变-裂变混合能源堆包括聚变中子源和以天然铀为燃料、水为冷却剂的次临界包层,主要目标是生产电力.利用输运燃耗耦合程序系统MCORGS计算了混合能源堆一维模型的燃耗,给出了中子有效增殖因数keff、能量放大倍数M、氚增殖比TBR等物理量随时间的变化.通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点.本文给出的结果可作为混合堆中子输运、燃耗分析程序校验的参考数据,为混合堆概念研究提供了基础数据.

  17. Investigation on using neutron counting techniques for online burnup monitoring of pebble bed reactor fuels

    Science.gov (United States)

    Zhao, Zhongxiang

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions by ORIGEN2. It was found that the neutron emission from an irradiated pebble increases with burnup super-linearly and reaches to 104 neutron/sec/pebble at the discharge burnup. The photon emission from an irradiated pebble was found to be in the order of 1013 photon/sec/pebble at all burnup levels. Analysis shows that the neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged one-group cross sections used in the depletion calculations, which consequently leads to large uncertainty in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and the neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting at low burnup levels. At high burnup levels, the uncertainty in the neutron emission rate becomes less but is still large in quantity. However, considering the super-linear feature of the correlation, the uncertainty in burnup determination was found to be ˜7% at the discharge burnup, which is acceptable. Therefore, total neutron emission rate of a pebble can be used as a burnup indicator to determine whether a pebble should be discharged or not. The feasibility of using passive neutron counting methods for the on-line burnup measurement was investigated by using a general Monte Carlo code, MCNP, to assess the detectability of the neutron emission and the capability to discriminate gamma noise by commonly used neutron detectors. It was found that both He-3

  18. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  19. Integrated burnup calculation code system SWAT

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Hirakawa, Naohiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; 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)

  20. 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.

  1. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    Energy Technology Data Exchange (ETDEWEB)

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  5. Energy corrections and persistent perturbation effects in continuous spectra

    NARCIS (Netherlands)

    Hove, Léon van

    1955-01-01

    The quantum-mechanical perturbation theory of continuous energy spectra is investigated for a special class of perturbations possessing some of the formal properties of the familiar interaction energies of field theory. These formal properties entail the inapplicability of the familiar perturbation

  6. Continuously Optimized Reliable Energy (CORE) Microgrid: Models & Tools (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    2013-07-01

    This brochure describes Continuously Optimized Reliable Energy (CORE), a trademarked process NREL employs to produce conceptual microgrid designs. This systems-based process enables designs to be optimized for economic value, energy surety, and sustainability. Capabilities NREL offers in support of microgrid design are explained.

  7. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  8. Bioconversion of energy by Spirulina maxima on continuous culture

    Energy Technology Data Exchange (ETDEWEB)

    Chaumont, D.; Gudin, C.; Delepine, R.; Asensi, A.

    1982-11-01

    The influence of the light intensity, photoperiod and residence time were studied on a continuous culture of Spirulina maxima. A maximum photosynthetic conversion yield of 13% on visible energy and a minimum quantic need of 23 photons per integrated CO/sub 2/ molecule were obtained in this way. By the application of continuous cultivation techniques a strain of Sprirulina developing on sea water was selected.

  9. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  10. Continuous Electron--Energy Variation of the Eindhoven Racetrack Microtron.

    Science.gov (United States)

    Theuws, W. H. C.; Botman, J. I. M.; Hagedoorn, H. L.

    1997-05-01

    Energy variation of the Eindhoven racetrack microtron, which has been designed as a fixed--energy electron accelerator at 75 MeV, is considered in this paper. By taking the orbit pattern in the RTM as constant and varying certain parameters continuous energy variation can be obtained. The microtron injector is a linac producing electrons between 6 and 12 MeV. The microtron cavity potential and the magnetic guide fields must be adapted to the injection energy in order to fulfil the synchronism condition. The transverse and longitudinal acceptance of the RTM are effected by deviations of the electron velocity from the speed of light, which are different for each parameter setting. An account of these effects is presented together with the energy--setting measurements by using one of the microtron magnets as a spectrometer.

  11. ENERGY RECOVERY FOR CONTINUOUS DYEING PROCESS IN TEXTILE INDUSTRY ENTERPRISES

    Directory of Open Access Journals (Sweden)

    V. N. Romaniuk

    2015-01-01

    Full Text Available The paper ascertains and presents alteration in the energy consumption as a consequence of utilizing the low-temperature waste streams commonly used in the lines of continuous dyeing at the finishing shops of textile enterprises of Belarus. The utilization realizes through the engagement of lithium-bromide absorption heat pumps with various energy characteristics such as the heating coefficient (relative conversion ratio COPhp = 1,15; 1,7; 2,2 and the heating capacity. The latter associates with the converted heat-flow energy utilization variant with the heat-transfer medium heating system scheme (one-, twoand multistage heating. The article considers transition to previously not applied service-water preheating due to the technological acceptance of feeding higher temperature water into the dyeing machine and widening specification of the heattransfer media. The authors adduce variants of internal and external energy use and their evaluation based on the relative energy and exergy characteristics. With results of the thermodynamic analysis of the modernized production effectiveness the researchers prove that alongside with traditional and apparent interior utilization of the energy associated with the stream heat recuperation, it is advisable to widen the range of applied heat-transfer media. The transition to the service water twoand multi-stage preheating is feasible. The study shows that the existing energy supply efficiency extremely low index-numbers improve by one or two degrees. Since they are conditioned, inter alia, by the machinery design, traditional approach to energy supply and heat-medium usage as well as the enterprise whole heating system answering requirements of the bygone era of cheap energy resources. The authors examine the continuous dyeing line modernization options intending considerable investments. Preliminary economic assessment of such inevitable modernization options for the enterprise entire heat-and-power system

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  13. Benchmarking of energy consumption of continuous galvanizing lines

    Science.gov (United States)

    Gopalakrishnan, B.; Chavan, R.

    2005-11-01

    A case study revealed that more than 13,500 MMBtu of energy is wasted annually when a single galvanizing line is off-production for hardware replacement for duration of a few hours every 2 weeks. This energy if utilized for production will yield about 13,000 tons of Galvanized Sheet Steel annually from a single galvanizing line. Thus for the 57 [1] hot dip galvanizing lines in US this figure results in a production loss of 741,000 tons/year. An attempt has been made to develop a spreadsheet that will take into account all the major energy consuming equipment in a typical hot dip continuous line. It maintains a track of the current production and energy consumption. It can simulate a scenario where either the number of shutdowns or the hours per shutdown will be reduced as a consequence of better material developed by the researchers. Different charts pertaining to energy consumed by different equipment group, total cost of energy spent on natural gas and electricity, MMBtu/Ton, Tons/Year and Production time before shutdowns assists the engineers decide the best operating stretch to suite their production rate and optimize energy consumption to some extent. Validation data gathered from the three well established galvanizing lines powers this spreadsheet to forecast annual increase in production and thus helps judge the performance of the new hardware.

  14. Machine Learning and Sensor Fusion for Estimating Continuous Energy Expenditure

    OpenAIRE

    Vyas, Nisarg; BodyMedia, Inc.; Farringdon, Jonathan; BodyMedia Inc.; Andre, David; Cerebellum Capital, Inc.; Stivoric, John Ivo; BodyMedia

    2012-01-01

    In this article we provide insight into the BodyMedia FIT armband system — a wearable multi-sensor technology that continuously monitors physiological events related to energy expenditure for weight management using machine learning and data modeling methods. Since becoming commercially available in 2001, more than half a million users have used the system to track their physiological parameters and to achieve their individual health goals including weight-loss. We describe several challenges...

  15. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

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

    Energy Technology Data Exchange (ETDEWEB)

    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{sub 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{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 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{sub 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{sub 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{sub 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

  17. Determination of deuterium–tritium critical burn-up parameter by four temperature theory

    Energy Technology Data Exchange (ETDEWEB)

    Nazirzadeh, M.; Ghasemizad, A. [Department of Physics, University of Guilan, 41335-1914 Rasht (Iran, Islamic Republic of); Khanbabei, B. [School of Physics, Damghan University, 36716-41167 Damghan (Iran, Islamic Republic of)

    2015-12-15

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  18. 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)

  19. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

  20. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.

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

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

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

  2. Burnup determination of a fuel element concerning different cooling times; Seguimiento del quemado de un elemento combustible, para diferentes tiempos de enfriamento

    Energy Technology Data Exchange (ETDEWEB)

    Henriquez, C.; Navarro, G.; Pereda, C.; Mutis, O. [Comision Chilena de Energia Nuclear, Santiago (Chile). Dept. de Aplicaciones Nucleares. Unidad de Reactores; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    In this work we report a complete set of measurements and some relevant results regarding the burnup process of a fuel element containing low enriched nuclear fuel. This fuel element was fabricated at the Plant of Fuel Elements of the Chilean Nuclear Energy Commission (CCHEN). Measurements were carried out using gamma-ray spectroscopy and the absolute burnup of the fuel element was determined. (author)

  3. Buoyancy Driven Mixing with Continuous Volumetric Energy Deposition

    Science.gov (United States)

    Wachtor, Adam J.; Jebrail, Farzaneh F.; Dennisen, Nicholas A.; Andrews, Malcolm J.; Gore, Robert A.

    2014-11-01

    An experiment involving a miscible fluid pair is presented which transitioned from a Rayleigh-Taylor (RT) stable to RT unstable configuration through continuous volumetric energy deposition (VED) by microwave radiation. Initially a light, low microwave absorbing fluid rested above a heavier, more absorbing fluid. The alignment of the density gradient with gravity made the system stable, and the Atwood number (At) for the initial setup was approximately -0.12. Exposing the fluid pair to microwave radiation preferentially heated the bottom fluid, and caused its density to drop due to thermal expansion. As heating of the bottom fluid continued, the At varied from negative to positive, and after the system passed through the neutral stability point, At = 0, buoyancy driven mixing ensued. Continuous VED caused the At to continue increasing and further drive the mixing process. Successful VED mixing required careful design of the fluid pair used in the experiment. Therefore, fluid selection is discussed, along with challenges and limitations of data collection using the experimental microwave facility. Experimental and model predictions of the neutral stability point, and onset of buoyancy driven mixing, are compared, and differences with classical, constant At RT driven turbulence are discussed.

  4. High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

    Science.gov (United States)

    Lemoine, F.

    1997-09-01

    Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning the transient fission gas behavior and the clad loading mechanisms. The importance of the rim zone is demonstrated based on three experiments resulting in clad failure at low enthalpy, which are explained by energetic considerations. High gas release in non-failure tests with low energy deposition underlines the importance of grain boundary and porosity gas. Measured final releases are strongly correlated to the microstructure evolution, depending on energy deposition, pulse width, initial and refabricated fuel rod design. Observed helium release can also increase internal pressure and gives hints to the gas behavior understanding.

  5. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  6. Optimal continuous-variable teleportation under energy constraint

    Science.gov (United States)

    Lee, Jaehak; Park, Jiyong; Nha, Hyunchul

    2017-05-01

    Quantum teleportation is one of the crucial protocols in quantum information processing. It is important to accomplish an efficient teleportation under practical conditions, aiming at a higher fidelity desirably using fewer resources. The continuous-variable (CV) version of quantum teleportation was first proposed using a Gaussian state as a quantum resource, while other attempts were also made to improve performance by applying non-Gaussian operations. We investigate the CV teleportation to find its ultimate fidelity under energy constraint identifying an optimal quantum state. For this purpose, we present a formalism to evaluate teleportation fidelity as an expectation value of an operator. Using this formalism, we prove that the optimal state must be a form of photon-number entangled states. We further show that Gaussian states are near optimal, while non-Gaussian states make a slight improvement and therefore are rigorously optimal, particularly in the low-energy regime.

  7. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Science.gov (United States)

    2013-11-12

    ...: U.S. Department of Energy, C/O Melissa Bates, 1955 Freemont Ave., MS 1235, Idaho Falls, ID 83415..., 1955 Fremont Ave., Attn: Melissa Bates, Idaho Falls, ID, between 8 a.m. and 3:30 p.m. MT, Monday.... Melissa Bates, Contracting Officers Representative, High Burnup Dry Storage Cask Research and...

  8. Transport and Burnup Numerical Simulation on the Liquid Blanket Burnup of In-Zinerater%In-Zinerater液态包层输运燃耗数值模拟

    Institute of Scientific and Technical Information of China (English)

    师学明; 杨俊云; 刘成安

    2014-01-01

    Z-Pinch惯性约束聚变是未来一种有竞争力的能源候选方案。Z-Pinch驱动的聚变裂变混合堆可高效地嬗变反应堆乏燃料中分离出的超铀元素。对美国Sandia国家实验室提出的In-Zinerater混合堆概念进行了中子学分析和数值模拟。在三维输运燃耗耦合程序MCORGS中增加了处理在线添加燃料与去除裂变产物的功能,实现了对液态燃料燃耗过程的模拟。增加6Li丰度和燃料初装量保持寿期初反应性不变,可以减缓寿期内反应性下降趋势。逐步增加包层内超铀元素装量,可以控制整个寿期内反应性基本恒定。聚变功率取20 MW,通过反应性控制,5年内包层能量放大倍数在160∼180之间,氚增殖比在1.5∼1.7之间,优于In-Zinerater基准设计方案。%Z-Pinch Inertial confinement fusion is a competitive candidate for future energy solution. A fusion-fission hybrid driven by Z-Pinch can be used to transmute transuranic elements from spent fuels of reactors efficiently. Analysis and numerical simulation of blanket neutronics of In-Zinerater, which is a fusion-fission hybrid concept design in Sandia National Laboratories, is given in this paper. Modification to the three dimension transport and burnup code MCORGS are done, so as to simulate continuous feeding and continuous chemical processing of the liquid fuel. Different combination of initial enrichment of 6Li and fuels loading in the blanket are selected to keep the same reactivity at begin of core. By this way, the decreasing trend of reactivity at life of the core can be lowered. The reactivity can be maintained constant by increasing the fuel loading in the core gradually as the burnup deepens. Given a 20 MW fusion power, by reactivity control, the blanket energy multiplication is around 160∼180 and tritium breed ratio 1.5∼1.7 in 5 years, which is a better result than Sandia’s original design.

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

    Science.gov (United States)

    Jo, Jungmin; Cheon, MunSeong; Kim, Jun Young; Rhee, T.; Kim, Junghee; Shi, Yue-Jiang; Isobe, M.; Ogawa, K.; Chung, Kyoung-Jae; Hwang, Y. S.

    2016-11-01

    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 3He 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.

  10. Continuous fiber ceramic composites for energy related applications. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-07

    The US Department of Energy has established the Continuous Fiber Ceramic Composites (CFCC) program to develop technology for the manufacture of CFCC`s for use in industrial applications where a reduction in energy usage or emissions could be realized. As part of this program, the Dow Chemical Company explored the manufacture of a fiber reinforced/self reinforced silicon nitride for use in industrial chemical processing. In Dow`s program, CFCC manufacturing technology was developed around traditional, cost effective, tape casting routes. Formulations were developed and coupled with unique processing procedures which enabled the manufacture of tubular green laminates of the dimension needed for the application. An evaluation of the effect of various fibers and fiber coatings on the properties of a fiber reinforced composites was also conducted. Results indicated that fiber coatings could provide composites exhibiting non-catastrophic failure and substantially improved toughness. However, an evaluation of these materials in industrial process environments showed that the material system chosen by Dow did not provide the required performance improvements to make replacement of current metallic components with CFCC components economically viable.

  11. Continuous fiber ceramic composites for energy related applications. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-07

    The US Department of Energy has established the Continuous Fiber Ceramic Composites (CFCC) program to develop technology for the manufacture of CFCC`s for use in industrial applications where a reduction in energy usage or emissions could be realized. As part of this program, the Dow Chemical Company explored the manufacture of a fiber reinforced/self reinforced silicon nitride for use in industrial chemical processing. In Dow`s program, CFCC manufacturing technology was developed around traditional, cost effective, tape casting routes. Formulations were developed and coupled with unique processing procedures which enabled the manufacture of tubular green laminates of the dimension needed for the application. An evaluation of the effect of various fibers and fiber coatings on the properties of a fiber reinforced composites was also conducted. Results indicated that fiber coatings could provide composites exhibiting non-catastrophic failure and substantially improved toughness. However, an evaluation of these materials in industrial process environments showed that the material system chosen by Dow did not provide the required performance improvements to make replacement of current metallic components with CFCC components economically viable.

  12. Calibration of burnup monitor in the Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oheda, K.; Naito, H.; Hirota, M. [Japan Nuclear Fuel Ltd., Aomori (Japan); Natsume, K. [Toshiba Corp., Yokohama, Kawasaki, Kanagawa (Japan); Kumanomido, H. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-07-01

    The Rokkasho Reprocessing Plant has adopted a credit for burnup in criticality control in the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. The burnup monitor system, prepared for BWR and PWR type fuel assemblies, nondestructively measures the burnup value and determines the residual U-235 enrichment in a spent fuel assembly, and criticality is controlled by the value of residual U-235 enrichment in SFSF and by the value of top 50 cm average burnup in the Dissolution Facility. The burnup monitor consists of three measurement systems; a Boss gamma-ray profile measurement system, a high resolution gamma-ray spectrometry system, and a passive neutron measurement system. The monitor sensitivity is calibrated against operator-declared burnup values through repetitive measurements of 100 spent fuel assemblies: BWR 8 X 8, PWR 14 X 14. and 17 X 17. The outline of the measurement methods, objectives of the calibration, actual calibration method, and an example of calibration performed in a demonstration experiment are presented. (author)

  13. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.

  14. Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marsodi E-mail: marsodi@batan.go.id; Lasman, K.A.S.; Nishihara, K. E-mail: nishi@omega.tokai.jaeri.go.jp; Osugi, T.; Tsujimoto, K.; Marsongkohadi; Su' ud, Z. E-mail: szaki@fi.itb.ac.id

    2002-12-01

    Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead-bismuth, Pb-Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.

  15. S∧4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    Science.gov (United States)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The S∧4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S∧4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S∧4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    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 {sup 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 {sup 233}U, {sup 242}Pu, {sup 150}Nd, and {sup 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.

  17. Effect of burn-up and high burn-up structure on spent nuclear fuel alteration

    Energy Technology Data Exchange (ETDEWEB)

    Clarens, F.; Gonzalez-Robles, E.; Gimenez, F. J.; Casas, I.; Pablo, J. de; Serrano, D.; Wegen, D.; Glatz, J. P.; Martinez-Esparza, A.

    2009-07-01

    In this report the results of the experimental work carried out within the collaboration project between ITU-ENRESA-UPC/CTM on spent fuel (SF) covering the period 2005-2007 were presented. Studies on both RN release (Fast Release Fraction and matrix dissolution rate) and secondary phase formation were carried out by static and flow through experiments. Experiments were focussed on the study of the effect of BU with two PWR SF irradiated in commercial reactors with mean burn-ups of 48 and 60 MWd/KgU and; the effect of High Burn-up Structure (HBS) using powdered samples prepared from different radial positions. Additionally, two synthetic leaching solutions, bicarbonate and granitic bentonite ground wa ter were used. Higher releases were determined for RN from SF samples prepared from the center in comparison with the fuel from the periphery. However, within the studied range, no BU effect was observed. After one year of contact time, secondary phases were observed in batch experiments, covering the SF surface. Part of the work was performed for the Project NF-PRO of the European Commission 6th Framework Programme under contract no 2389. (Author)

  18. An empirical formulation to describe the evolution of the high burnup structure

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-15

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  19. An empirical formulation to describe the evolution of the high burnup structure

    Science.gov (United States)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-01

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  20. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  1. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

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

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

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

  3. Quantum Energy Teleportation with Electromagnetic Field: Discrete vs. Continuous Variables

    CERN Document Server

    Hotta, Masahiro

    2009-01-01

    Local measurements of quantum fluctuation in the vacuum state of electromagnetic field require energy infusion to the field. The infused energy is diffused to spatial infinity with light velocity and the state of the field soon becomes a local vacuum with zero energy around the measurement area. Of cource we cannot retrieve energy from this measurement area if we do not know the measurement result of the fluctuation. However, if the measurement result is available for us, we are able to extract energy from the local vacuum of the field, applying the protocol of quantum energy teleportation recently proposed. By performing a local unitary operation around the measurement area dependent on the measurement result, the fluctuaion of zero-point oscillation is squeezed and negative energy density appears around the area, accompanied by extraction of positive energy from the field. In this paper, we compare two different protocols of the energy retrieval. In the first protocol, a 1/2 spin is coupled with the fluctua...

  4. FY14 Status Report: CIRFT Testing Results on High Burnup UNF

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

    The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4, and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.

  5. Power excursion analysis for BWR`s at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Neymoith, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  6. Detailed Burnup Calculations for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, F. [Centro Atomico Bariloche (CNEA), 8400 S. C. de Bariloche (Argentina)

    2011-07-01

    A general method (RRMCQ) has been developed by introducing a microscopic burn up scheme which uses the Monte Carlo calculated spatial power distribution of a research reactor core and a depletion code for burn up calculations, as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy dependent cross-section libraries and full 3D geometry of the system is input for the calculations. The resulting predictions for the system at successive burn up time steps are thus based on a calculation route where both geometry and cross-sections are accurately represented, without geometry simplifications and with continuous energy data. The main advantage of this method over the classical deterministic methods currently used is that RRMCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burn up codes adopted until now are the widely used MCNP5 and ORIGEN2 codes, but other codes can be used also. For using this method, there is a need of a well-known set of nuclear data for isotopes involved in burn up chains, including burnable poisons, fission products and actinides. For fixing the data to be included on this set, a study of the present status of nuclear data is performed, as part of the development of RRMCQ method. This study begins with a review of the available cross-section data of isotopes involved in burn up chains for research nuclear reactors. The main data needs for burn up calculations are neutron cross-sections, decay constants, branching ratios, fission energy and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross-section data for burn up calculations, using some of the main available evaluated nuclear data files. Basically, the RRMCQ detailed burn up method includes four

  7. Understanding change and continuity in residential energy consumption

    DEFF Research Database (Denmark)

    Gram-Hanssen, Kirsten

    2011-01-01

    of material consumer goods in practice theory. Case studies on household energy consumption are used as an empirical basis for these discussions. Looking at household energy consumption through the theoretical lens of practice theory necessitates discussion on whether energy consumption should be viewed......Practice theory has recently emerged within consumer studies as a promising approach that shifts focus from the individual consumer towards the collective aspects of consumption and from spectacular and conspicuous dimensions of consumption towards routine and mundane aspects of consumption...

  8. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  9. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  10. Energy balance during two days of continuous stationary cycling

    Directory of Open Access Journals (Sweden)

    Stewart Kelly L

    2007-10-01

    Full Text Available Abstract This study examined the capabilities of an ultraendurance athlete to self-regulate their diet during an attempt on the record for the longest period of stationary cycling. The attempt required the athlete to complete at least 20 km/hr, with a 15 minute break allowed every eight hours. Laboratory tests determined a heart rate-oxygen consumption regression equation enabling calculation of energy expenditure from heart rate during the attempt. Energy intake was determined by a non-weighed dietary record collected at the time of consumption. The athlete completed 46.7 hours, covering 1126 km, at a speed of 24 ± 1.6 km/hr. He expended 14486 kcal and consumed 11098 kcal resulting in an energy deficit (-3290 kcal and a weight loss (-0.55 kg. The carbohydrate (42 ± 32 g/hr, water (422 ± 441 ml/hr, and sodium (306 ± 465 mg/hr intake were all below current recommendations. The athlete was unable to self-regulate his diet or exercise intensity to prevent a negative energy balance.

  11. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

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

  13. Energy Continuity in Degenerate Density Functional Perturbation Theory

    CERN Document Server

    Palenik, Mark C

    2016-01-01

    Fractional occupation numbers can produce open-shell degeneracy in density functional theory. We develop the corresponding perturbation theory by requiring that a differentiable map connects the initial and perturbed states. The degenerate state connects to a single perturbed state which extremizes, but does not necessarily minimize or maximize, the energy with respect to occupation numbers. Using a system of three electrons in a harmonic oscillator potential, we relate the counterintuitive sign of first-order occupation numbers to eigenvalues of the electron-electron interaction Hessian.

  14. Proposed continuous wave energy recovery operation of an XFEL

    Energy Technology Data Exchange (ETDEWEB)

    J. Sekutowicz; S. A. Bogacz; D. Douglas; P. Kneisel; G. P. Williams; M. Ferrario; L. Serafini; I. Ben-Zvi; J. Rose; J. Smedley; T. Srinivasan-Rao; W.-D. Moeller; B. Petersen; D. Proch; S. Simrock; P. Colestock; J. B. Rosenzweig

    2004-05-01

    Commissioning of two large coherent light facilities at SLAC and DESY should begin in 2008 and in 2011 respectively. In this paper we look further into the future, hoping to answer, in a very preliminary way, two questions. First: ''What will the next generation of XFEL facilities look like?'' Believing that superconducting technology offers advantages such as high quality beams with highly populated bunches, the possibility of energy recovery and higher overall efficiency than warm technology, we focus this preliminary study on the superconducting option. From this belief the second question arises: ''What modifications in superconducting technology and in the machine design are needed, as compared to the present DESY XFEL, and what kind of R&D program should be proposed to arrive in the next few years at a technically feasible solution with even higher brilliance and increased overall conversion of AC power to photon beam power?'' In this paper we will very often refer to and profit from the DESY XFEL design, acknowledging its many technically innovative solutions.

  15. Continuous wave energy recovery operation of an XFEL

    Energy Technology Data Exchange (ETDEWEB)

    Jacek Sekutowicz; S. A. Bogacz; D. Douglas; Peter Kneisel; G. P. Williams; M. Ferrario; L. Serafini; I. Ben-Zvi; J. Rose; T. Srinivasan-Rao; W.-D. Mueller; B. Petersen; D. Proch; S.Simrock; P. Colestock; J. B. Rosenzweig

    2003-12-01

    Commissioning of two large coherent light facilities at SLAC and DESY should begin in 2008 and in 2011 respectively. In this paper we look further into the future, hoping to answer, in a very preliminary way, two questions. First: ''What will the next generation of XFEL facilities look like?'' Believing that superconducting technology offers advantages such as high quality beams with highly populated bunches, the possibility of energy recovery and higher overall efficiency than warm technology, we focus this preliminary study on the superconducting option. From this belief the second question arises: ''What modifications in superconducting technology and in the machine design are needed, as compared to the present DESY XFEL, and what kind of R&D program should be proposed to arrive in the next few years at a technically feasible solution with even higher brilliance and increased overall conversion of AC power to photon beam power?'' In this paper we will very often refer to and profit from the DESY XFEL design, acknowledging its many technically innovative solutions.

  16. OREST - The hammer-origen burnup program system

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U. (Gesellschaft fur Reaktorsicherheit mbH Forschungsgelande, 8046 Garching bei Munchen (DE))

    1988-08-01

    Reliable prediction of the characteristics of irradiated light water reactor fuels (e.g., afterheat power, neutron and gamma radiation sources, final uranium and plutonium contents) is needed for many aspects of the nuclear fuel cycle. Two main problems must be solved: the simulation of all isotopic nuclear reactions and the simulation of neutron fluxes setting the reactions in motion. In state-of-the-art computer techniques, a combination of specialized codes for lattice cell and burnup calculations is preferred to solve these cross-linked problems in time or burnup step approximation. In the program system OREST, developed for official and commercial tasks in the Federal Republic of Germany nuclear fuel cycle, the well-known codes HAMMER and ORIGEN and directly coupled with a fuel rod temperature module.

  17. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  18. A high burnup model developed for the DIONISIO code

    Energy Technology Data Exchange (ETDEWEB)

    Soba, A. [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, A., E-mail: denis@cnea.gov.ar [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Romero, L. [U.A. Reactores Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Villarino, E.; Sardella, F. [Departamento Ingeniería Nuclear, INVAP SE, Comandante Luis Piedra Buena 4950, 8430 San Carlos de Bariloche, Río Negro (Argentina)

    2013-02-15

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO{sub 2} fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in {sup 235}U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  19. A high burnup model developed for the DIONISIO code

    Science.gov (United States)

    Soba, A.; Denis, A.; Romero, L.; Villarino, E.; Sardella, F.

    2013-02-01

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  20. Tritium Burn-up Depth and Tritium Break-Even Time

    Institute of Scientific and Technical Information of China (English)

    LI Cheng-Yue; DENG Bai-Quan; HUANG Jin-Hua; YAN Jian-Cheng

    2006-01-01

    @@ Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder (FEB). It is found that the tritium burn-up depth is 317g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18kg that is more than theleast required amount of tritium storage to start up three of FEB-like fusion reactors.

  1. A multi-platform linking code for fuel burnup and radiotoxicity analysis

    Science.gov (United States)

    Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

    2014-02-01

    A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

  2. Modeling of burnup express-estimation for UO{sub 2}-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Likhanskii, Vladimir V.; Tokarev, Sergey A.; Vilkhivskaya, Olga V., E-mail: vilhivskaya_olga@mail.ru

    2017-03-15

    Highlights: • Proposed engineering model estimates fuel burnup by {sup 134}Cs/{sup 137}Cs activity ratio. • Buildup of cesium isotopes relies on changing neutron spectrum in the core cycle. • {sup 134}Cs/{sup 137}Cs activity ratios in FAs with Gd-doped fuel rods are analyzed. • Comparison of the model calculations with the NPPs spike measurements is presented. - Abstract: The paper presents the developed engineering model of cesium isotopes production as function of UO{sub 2}-fuel burnup and an assessment of their activity ratios. The model considers the evolution of linear power of gadolinium-doped fuel rods and fuel rods surrounding them in fuel assemblies with high enrichment fuel, harder neutron spectrum, and the changes in cross-sections of neutron reactions in thermal and epithermal energy areas. Parametrical dependences in the model are based on the fuel operation data for nuclear power plants and on the detailed neutronic-physical calculations of the core. Presented are the results of the model calculations for the {sup 134}Cs/{sup 137}Cs activity ratios in fuel taking into account the parameter of hardness of the neutron spectrum during the first irradiation cycle for fuel with enrichment ranging from 3.6 wt% in {sup 235}U.

  3. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    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)

  4. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  5. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  6. The continuous tower of scalar fields as a system of interacting dark matter–dark energy

    Directory of Open Access Journals (Sweden)

    Paulo Santos

    2015-10-01

    Full Text Available This paper aims to introduce a new parameterisation for the coupling Q in interacting dark matter and dark energy models by connecting said models with the Continuous Tower of Scalar Fields model. Based upon the existence of a dark matter and a dark energy sectors in the Continuous Tower of Scalar Fields, a simplification is considered for the evolution of a single scalar field from the tower, validated in this paper. This allows for the results obtained with the Continuous Tower of Scalar Fields model to match those of an interacting dark matter–dark energy system, considering that the energy transferred from one fluid to the other is given by the energy of the scalar fields that start oscillating at a given time, rather than considering that the energy transference depends on properties of the whole fluids that are interacting.

  7. 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)

  8. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    Energy Technology Data Exchange (ETDEWEB)

    Leal, Luiz C [ORNL; Derrien, Herve [ORNL; Dunn, Michael E [ORNL; Mueller, Don [ORNL

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  9. 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

  10. Energy-like conserved quantity of a nonlinear nonconsevative continuous system

    Institute of Scientific and Technical Information of China (English)

    CHEN Liqun

    2004-01-01

    A system whose energy is not conserved is called nonconservative. To investigate if there exists a conserved quantity that has the same dimension as energy and is positively definite, the author analyzed the bending vibration of an axially moving beam with geometric nonlinearity.Based on the governing equation, the energy was proven to be not conserved in the case where the beam has two simply supported or fixed ends. A definitely positive quantity with the energy dimension was defined. The quantity was verified to remain a constant during the motion. The investigation indicates that an energy-like conserved quantity may exist in a nonlinear nonconservative continuous system.

  11. Continuous sterilization process developed for offal processing needs half the amount of energy necessary for batch processing. Continu sterilisatieproces voor slachtafval vergt helft minder energie

    Energy Technology Data Exchange (ETDEWEB)

    Walraven, O.E.D. (Stork Duke, Boxmeer (Netherlands))

    1990-09-01

    Offal is processed into meat flour and fat by dehydration. During this process the offal has to be sterilized. This is normally done by batch processing. Stork Duke has developed a continuous sterilization process with financial support from the European Communities. As a result of less energy consumption and operational advantages the extra investments have payback periods from three to four years. Additional advantages are better product quality and reduced odor emission. 2 figs., 4 refs., 2 ills.

  12. Tight Uniform Continuity Bounds for Quantum Entropies: Conditional Entropy, Relative Entropy Distance and Energy Constraints

    Science.gov (United States)

    Winter, Andreas

    2016-10-01

    We present a bouquet of continuity bounds for quantum entropies, falling broadly into two classes: first, a tight analysis of the Alicki-Fannes continuity bounds for the conditional von Neumann entropy, reaching almost the best possible form that depends only on the system dimension and the trace distance of the states. Almost the same proof can be used to derive similar continuity bounds for the relative entropy distance from a convex set of states or positive operators. As applications, we give new proofs, with tighter bounds, of the asymptotic continuity of the relative entropy of entanglement, E R , and its regularization {E_R^{∞}}, as well as of the entanglement of formation, E F . Using a novel "quantum coupling" of density operators, which may be of independent interest, we extend the latter to an asymptotic continuity bound for the regularized entanglement of formation, aka entanglement cost, {E_C=E_F^{∞}}. Second, we derive analogous continuity bounds for the von Neumann entropy and conditional entropy in infinite dimensional systems under an energy constraint, most importantly systems of multiple quantum harmonic oscillators. While without an energy bound the entropy is discontinuous, it is well-known to be continuous on states of bounded energy. However, a quantitative statement to that effect seems not to have been known. Here, under some regularity assumptions on the Hamiltonian, we find that, quite intuitively, the Gibbs entropy at the given energy roughly takes the role of the Hilbert space dimension in the finite-dimensional Fannes inequality.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  14. Analysis of Discharge Spark Energy in Buck Converter of a Continuous Mode of Inductive Current

    Institute of Scientific and Technical Information of China (English)

    CUI Bao-chun; CHENG Hong; WANG Cong; LU Huan-yu; SHI Yun

    2006-01-01

    The basic idea of intrinsically safe circuit and the discharge spark in the Buck converter in the explosive atmospheres were introduced. The Buck converter is the main topological structure of the switch type of intrinsically safe circuit, which has two working modes: continuous inductive current (CCM - continuous conduction mode) and discrete inductance current (DCM - discontinuous conduction mode). The operating state of the continuous inductive current mode is analyzed in detail and the energy of discharge spark in various operating modes is discussed. The total energy will decrease with the increase of switch frequency, in a switching cycle; the discharge spark energy has a maximum and a minimum value. Therefore, the Buck converter has smaller discharge spark energy than the linear power circuit and the switch type of intrinsically safe circuit can enhance the output power and the conversion efficiency of the intrinsically safe power.

  15. A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

    2011-05-01

    A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

  16. MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, T.; Sternat, M.; Charlton, W.

    2011-05-08

    MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-19

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

  18. Determination of IRT-2M fuel burnup by gamma spectrometry.

    Science.gov (United States)

    Koleška, Michal; Viererbl, Ladislav; Marek, Milan; Ernest, Jaroslav; Šunka, Michal; Vinš, Miroslav

    2016-01-01

    A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as (137)Cs, (134)Cs, (144)Ce, (106)Ru and (154)Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes.

  19. Thermodynamic analysis for high burn-up fuel internal chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan). Faculty of Science and Technology

    1997-09-01

    Chemical states of fission products and actinide elements in high burn-up LWR fuel pellets have been analyzed thermodynamically using the computer program SOLGASMIX-PV. Calculations with this computer code have been performed for a complex multi-component system, which comprises 54 chemical species. The analysis shows that neither alkali nor alkaline-earth uranates are formed, but alkali and alkaline-earth molybdates exist in irradiated LWR fuel pellets in contrast with their post irradiation examinations. These molybdates tend to increase with increasing oxygen potential in the fuel under operating conditions, whereas the zirconates decrease. (author)

  20. Fabrication characteristics of dry process fuel with a variation of fuel burn-ups

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Kim, W. K.; Lee, J. W. [and others

    2004-11-01

    Fabrication characteristics of the dry processed fuel with a variation of fuel burn-ups in a range of 27,300 to 65,000 MWD/tU were experimentally evaluated. Density comparison of powders which were fabricated from oxidation, OREOX and milling processes at same process conditions was performed with a function of fuel burn-ups respectively. The influence of fuel burn-ups on sintering characteristics of dry processed fuel was studied by comparing the density change of sintered pellet as well as green pellet. Weight gain by fuel oxidation to U{sub 3}O{sub 8} showed semi-linear dependence with increasing fuel burn-ups. OREOX powder density increased up to 3.7 g/cm{sup 3} at high burn-up fuel, and the density of milled powder with fuel burn-ups represented almost similar value of 3.2{+-}0.2 g/cm{sup 3}. Also, the green pellet density compacted by 120 MPa decreased smoothly with increasing fuel burn-ups, and the density change of sintered pellet showed the similar trend as green pellet. The sintered density of pellet in a range of 27,000 to 40,000 MWD/tU was found to be more 95% of Theoretical Density(T.D.), but the sintered pellet density fabricated from high burn-up fuel showed a range of 92 % to 93% of T.D.

  1. 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)

  2. 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.

  3. High burnup effects on fuel behaviour under accident conditions: the tests CABRI REP-Na

    Science.gov (United States)

    Schmitz, Franz; Papin, Joelle

    A large, performance based, knowledge and experience in the field of nuclear fuel behaviour is available for nominal operation conditions. The database is continuously completed and precursor assembly irradiations are performed for testing of new materials and innovative designs. This procedure produces data and arguments to extend licencing limits in the permanent research for economic competitiveness. A similar effort must be devoted to the establishment of a database for fuel behaviour under off-normal and accident conditions. In particular, special attention must be given to the so-called design-basis-accident (DBA) conditions. Safety criteria are formulated for these situations and must be respected without consideration of the occurrence probability and the risk associated to the accident situation. The introduction of MOX fuel into the cores of light water reactors and the steadily increasing goal burnup of the fuel call for research work, both experimental and analytical, in the field of fuel response to DBA conditions. In 1992, a significant programme step, CABRI REP-Na, has been launched by the French Nuclear Safety and Protection Institute (IPSN) in the field of the reactivity initiated accident (RIA). After performing the nine experiments of the initial test matrix it can be concluded that important new findings have been evidenced. High burnup clad corrosion and the associated degradation of the mechanical properties of the ZIRCALOY4 clad is one of the key phenomena of the fuel behaviour under accident conditions. Equally important is the evidence that transient, dynamic fission gas effects resulting from the close to adiabatic heating introduces a new explosive loading mechanism which may lead to clad rupture under RIA conditions, especially in the case of heterogeneous MOX fuel.

  4. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

  5. 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.

  6. 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)

  7. CONTINUOUS-ENERGY MONTE CARLO METHODS FOR CALCULATING GENERALIZED RESPONSE SENSITIVITIES USING TSUNAMI-3D

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, Christopher M [ORNL; Rearden, Bradley T [ORNL

    2014-01-01

    This work introduces a new approach for calculating sensitivity coefficients for generalized neutronic responses to nuclear data uncertainties using continuous-energy Monte Carlo methods. The approach presented in this paper, known as the GEAR-MC method, allows for the calculation of generalized sensitivity coefficients for multiple responses in a single Monte Carlo calculation with no nuclear data perturbations or knowledge of nuclear covariance data. The theory behind the GEAR-MC method is presented here, and proof of principle is demonstrated by using the GEAR-MC method to calculate sensitivity coefficients for responses in several 3D, continuous-energy Monte Carlo applications.

  8. Continuous-energy eigenvalue sensitivity coefficient calculations in TSUNAMI-3D

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, C. M.; Rearden, B. T. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Division, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)

    2013-07-01

    Two methods for calculating eigenvalue sensitivity coefficients in continuous-energy Monte Carlo applications were implemented in the KENO code within the SCALE code package. The methods were used to calculate sensitivity coefficients for several test problems and produced sensitivity coefficients that agreed well with both reference sensitivities and multigroup TSUNAMI-3D sensitivity coefficients. The newly developed CLUTCH method was observed to produce sensitivity coefficients with high figures of merit and a low memory footprint, and both continuous-energy sensitivity methods met or exceeded the accuracy of the multigroup TSUNAMI-3D calculations. (authors)

  9. Development of a SCALE Tool for Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, Christopher M [ORNL; Rearden, Bradley T [ORNL

    2013-01-01

    Two methods for calculating eigenvalue sensitivity coefficients in continuous-energy Monte Carlo applications were implemented in the KENO code within the SCALE code package. The methods were used to calculate sensitivity coefficients for several criticality safety problems and produced sensitivity coefficients that agreed well with both reference sensitivities and multigroup TSUNAMI-3D sensitivity coefficients. The newly developed CLUTCH method was observed to produce sensitivity coefficients with high figures of merit and low memory requirements, and both continuous-energy sensitivity methods met or exceeded the accuracy of the multigroup TSUNAMI-3D calculations.

  10. Rare earths & climate change,new energy,energy conservation and pollution reduction(continued)

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    @@ Ⅲ.Contribution of rare earths to energy conservation Rechargeable batteries and rare earth permanent magnetic motor matching with batteries in every Prius car consume approximately 10 kg of rare-earth hydro-gen storage materials and 2 kg of rare earth permanent magnetic materials respectively.

  11. Extension and validation of the TRANSURANUS burn-up model for helium production in high burn-up LWR fuels

    Science.gov (United States)

    Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang

    2011-12-01

    The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238-242Pu, 241Am, 243Am and 242-245Cm isotopes are described. Experimental data used for the extended validation include new EPMA measurements of the local concentrations of Nd and Pu and recent SIMS measurements of the radial distributions of Pu, Am and Cm isotopes, both in a 3.5% enriched commercial PWR UO 2 fuel with a burn-up of 80 and 65 MWd/kgHM, respectively. Good agreement has been found between TUBRNP and the experimental data. The analysis has been complemented by detailed neutron transport calculations (VESTA code), and also revealed the need to update the branching ratio for the 241Am(n,γ) 242mAm reaction in typical PWR conditions.

  12. Energy control in the Palais des Sports et des Congres in Megeve: a continuous step

    Energy Technology Data Exchange (ETDEWEB)

    Tourneur, H. [Palais des Sports et des Congres, 74 - Megeve (France)

    1993-12-31

    Energy control in a cultural and sport complex is constantly improvable, if the care is continuous. It requires analysis, organization, watchfulness, awareness, intervention versatility and, in every moment, calling into question of systems which have been arranged according to changing data. This will is existing in Megeve`s Palais des Sports. The results are speaking for themselves. (Author).

  13. Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

    Energy Technology Data Exchange (ETDEWEB)

    T. Downar

    2009-03-31

    The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system.

  14. Evaluation of an integrated continuous stirred microbial electrochemical reactor: Wastewater treatment, energy recovery and microbial community.

    Science.gov (United States)

    Wang, Haiman; Qu, Youpeng; Li, Da; Zhou, Xiangtong; Feng, Yujie

    2015-11-01

    A continuous stirred microbial electrochemical reactor (CSMER) was developed by integrating anaerobic digestion (AD) and microbial electrochemical system (MES). The system was capable of treating high strength artificial wastewater and simultaneously recovering electric and methane energy. Maximum power density of 583±9, 562±7, 533±10 and 572±6 mW m(-2) were obtained by each cell in a four-independent circuit mode operation at an OLR of 12 kg COD m(-3) d(-1). COD removal and energy recovery efficiency were 87.1% and 32.1%, which were 1.6 and 2.5 times higher than that of a continuous stirred tank reactor (CSTR). Larger amount of Deltaproteobacteria (5.3%) and hydrogenotrophic methanogens (47%) can account for the better performance of CSMER, since syntrophic associations among them provided more degradation pathways compared to the CSTR. Results demonstrate the CSMER holds great promise for efficient wastewater treatment and energy recovery.

  15. Energy of hydrodynamic and magnetohydrodynamic waves with point and continuous spectra

    Science.gov (United States)

    Hirota, M.; Fukumoto, Y.

    2008-08-01

    Energy of waves (or eigenmodes) in an ideal fluid and plasma is formulated in the noncanonical Hamiltonian context. By imposing the kinematical constraint on perturbations, the linearized Hamiltonian equation provides a formal definition of wave energy not only for eigenmodes corresponding to point spectra but also for singular ones corresponding to a continuous spectrum. The latter becomes dominant when mean fields have inhomogeneity originating from shear or gradient of the fields. The energy of each wave is represented by the eigenfrequency multiplied by the wave action, which is nothing but the action variable and, moreover, is associated with a derivative of a suitably defined dispersion relation. The sign of the action variable is crucial to the occurrence of Hopf bifurcation in Hamiltonian systems of finite degrees of freedom [M. G. Krein, Dokl. Akad. Nauk SSSR, Ser. A 73, 445 (1950)]. Krein's idea is extended to the case of coalescence between point and continuous spectra.

  16. 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.

  17. 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).

  18. Burnup calculations for the HOMER-15 and SAFE-300 reactors

    Science.gov (United States)

    Amiri, Benjamin W.; Poston, David I.

    2002-01-01

    The Heatpipe Power System (HPS) is a near-term low-cost space fission power system. As the U-235 fuel of the HPS is burned, higher actinides and fission products will be produced. This will cause changes in system reactivity, radioactivity, and decay power. One potential concern is that gaseous fission products may exert excessive pressure on the fuel pin cladding. To evaluate these issues, simulations were run in MONTEBURNS. MONTEBURNS is an automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. This paper describes the results of these simulations, as well as how those results compare with the current experimental database of irradiated materials. .

  19. 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.

  20. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  1. Impact energy analysis of turbulent water sprays for continuous centrifugal concentration

    Institute of Scientific and Technical Information of China (English)

    PEN Nan-qi; CHEN Lu-zheng; XIONG Da-he

    2009-01-01

    A SLon full-scale continuous centrifugal concentrator was used to reconcentrate hematite from a high gradient magnetic separation concentrate to study the effect of impact angle, concentrate mass and drum rotation speed on the impact energy of turbulent water sprays for continuous centrifugal concentration, under conditions of feed volume flow rate around 9 m3/h, feed solid concentration of 25% - 35% and reciprocating velocity of water sprays at 0. 05 m/s. The results indicate that a mimmal critical impact energy is required in the water sprays for achieving continuous concentration of the concentrator; an unfitted impact angle reduces the impact ffciency, and the highest impact efficiency of 0. 6416 is found at the mpact angle of 60°; the increase in concentrate mass leads to an increase in impact energy, and the highest impact efficiency is maintained when the concentrate mass varies in the range of 0. 44 -0. 59 kg/s; when the concentrate mass and the pressure of water sprays are kept at around 0. 45 kg/s and in the range of 0. 4 -0. 6 MPa respectively, the impact energy increases proportionally with the increase of drum rotation speed.

  2. 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)

  3. 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

  4. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  5. Estimating the continuous-time dynamics of energy and fat metabolism in mice.

    Directory of Open Access Journals (Sweden)

    Juen Guo

    2009-09-01

    Full Text Available The mouse has become the most popular organism for investigating molecular mechanisms of body weight regulation. But understanding the physiological context by which a molecule exerts its effect on body weight requires knowledge of energy intake, energy expenditure, and fuel selection. Furthermore, measurements of these variables made at an isolated time point cannot explain why body weight has its present value since body weight is determined by the past history of energy and macronutrient imbalance. While food intake and body weight changes can be frequently measured over several weeks (the relevant time scale for mice, correspondingly frequent measurements of energy expenditure and fuel selection are not currently feasible. To address this issue, we developed a mathematical method based on the law of energy conservation that uses the measured time course of body weight and food intake to estimate the underlying continuous-time dynamics of energy output and net fat oxidation. We applied our methodology to male C57BL/6 mice consuming various ad libitum diets during weight gain and loss over several weeks and present the first continuous-time estimates of energy output and net fat oxidation rates underlying the observed body composition changes. We show that transient energy and fat imbalances in the first several days following a diet switch can account for a significant fraction of the total body weight change. We also discovered a time-invariant curve relating body fat and fat-free masses in male C57BL/6 mice, and the shape of this curve determines how diet, fuel selection, and body composition are interrelated.

  6. POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, M.E.; Greene, N.M.

    2000-12-01

    POLIDENT (Point Libraries of Data from ENDF/B Tapes) is an AMPX module that accesses the resonance parameters from File 2 of an ENDF/B library and constructs the continuous-energy cross sections in the resonance energy region. The cross sections in the resonance range are subsequently combined with the File 3 background data to construct the cross-section representation over the complete energy range. POLIDENT has the capability to process all resonance reactions that are identified in File 2 of the ENDF/B library. In addition, the code has the capability to process the single- and multi-level Breit-Wigner, Reich-Moore and Adler-Adler resonance formalisms that are identified in File 2. POLIDENT uses a robust energy-mesh-generation scheme that determines the minimum, maximum and points of inflection in the cross-section function in the resolved-resonance region. Furthermore, POLIDENT processes all continuous-energy cross-section reactions that are identified in File 3 of the ENDF/B library and outputs all reactions in an ENDF/B TAB1 format that can be accessed by other AMPX modules.

  7. POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, M.E.

    2000-10-20

    POLIDENT (POint LIbraries of Data from ENDF/B Tapes) is an AMPX module that accesses the resonance parameters from File 2 of an ENDF/B library and constructs the continuous-energy cross sections in the resonance energy region. The cross sections in the resonance range are subsequently combined with the File 3 background data to construct the cross-section representation over the complete energy range. POLIDENT has the capability to process all resonance reactions that are identified in File 2 of the ENDF/B library. In addition, the code has the capability to process the single- and multi-level Breit-Wigner, Reich-Moore and Adler-Adler resonance formalisms that are identified in File 2. POLIDENT uses a robust energy-mesh-generation scheme that determines the minimum, maximum and points of inflection in the cross-section function in the resolved-resonance region. Furthermore, POLIDENT processes all continuous-energy cross-section reactions that are identified in File 3 of the ENDF/B library and outputs all reactions in an ENDF/B TAB1 format that can be accessed by other AMPX modules.

  8. A power conditioning system with kinetic energy storage using a continuously variable gearbox

    Energy Technology Data Exchange (ETDEWEB)

    Ruddell, A.J.; Freris, L.L. [Imperial Coll. of Science, Technology and Medicine, London (United Kingdom); Bleijs, J.A.M. [Leicester Univ. (United Kingdom); Infield, D.G. [Rutherford Appleton Lab., Chilton (United Kingdom)

    1995-12-31

    Integration of wind power into isolated networks and weak mains grids presents problems due to power fluctuations. Energy storage may be used to compensate for wind power fluctuations at the source. This paper discusses the requirements of an energy store, and describes details of a prototype system consisting of a flywheel, a mechanical continuously variable transmission (CVT), and a synchronous machine. The requirements for control of power, voltage, and frequency of the system are defined. Computer models and simulation are used to design suitable controllers, and optimise the dynamic response of the power conditioning system. The prototype system has been constructed and commissioned and results of initial tests are described. (Author)

  9. Continuous thermal hydrolysis and energy integration in sludge anaerobic digestion plants.

    Science.gov (United States)

    Fdz-Polanco, F; Velazquez, R; Perez-Elvira, S I; Casas, C; del Barrio, D; Cantero, F J; Fdz-Polanco, M; Rodriguez, P; Panizo, L; Serrat, J; Rouge, P

    2008-01-01

    A thermal hydrolysis pilot plant with direct steam injection heating was designed and constructed. In a first period the equipment was operated in batch to verify the effect of sludge type, pressure and temperature, residence time and solids concentration. Optimal operation conditions were reached for secondary sludge at 170 degrees C, 7 bar and 30 minutes residence time, obtaining a disintegration factor higher than 10, methane production increase by 50% and easy centrifugation In a second period the pilot plant was operated working with continuous feed, testing the efficiency by using two continuous anaerobic digester operating in the mesophilic and thermophilic range. Working at 12 days residence time, biogas production increases by 40-50%. Integrating the energy transfer it is possible to design a self-sufficient system that takes advantage of this methane increase to produce 40% more electric energy.

  10. Development of Continuous-Energy Eigenvalue Sensitivity Coefficient Calculation Methods in the Shift Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, Christopher M [ORNL; Martin, William R [University of Michigan; Rearden, Bradley T [ORNL; Williams, Mark L [ORNL

    2012-01-01

    Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the SHIFT Monte Carlo code within the Scale code package. The methods were used for several simple test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods.

  11. Use of SCALE Continuous-Energy Monte Carlo Tools for Eigenvalue Sensitivity Coefficient Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, Christopher M [ORNL; Rearden, Bradley T [ORNL

    2013-01-01

    The TSUNAMI code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, such as quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the development of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The CLUTCH and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in the CE KENO framework to generate the capability for TSUNAMI-3D to perform eigenvalue sensitivity calculations in continuous-energy applications. This work explores the improvements in accuracy that can be gained in eigenvalue and eigenvalue sensitivity calculations through the use of the SCALE CE KENO and CE TSUNAMI continuous-energy Monte Carlo tools as compared to multigroup tools. The CE KENO and CE TSUNAMI tools were used to analyze two difficult models of critical benchmarks, and produced eigenvalue and eigenvalue sensitivity coefficient results that showed a marked improvement in accuracy. The CLUTCH sensitivity method in particular excelled in terms of efficiency and computational memory requirements.

  12. 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.

  13. Fission-gas release at extended burnups: effect of two-dimensional heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Yu, S.D. [Ryerson Polytechnic Univ., Toronto, Ontario (Canada); Lau, J.H.K

    2000-09-01

    To better simulate the performance of high-burnup CANDU fuel, a two-dimensional model for heat transfer between the pellet and the sheath has been added to the computer code ELESTRES. The model covers four relative orientations of the pellet and the sheath and their impacts on heat transfer and fission-gas release. The predictions of the code were compared to a database of 27 experimental irradiations involving extended burnups and normal burnups. The calculated values of fission gas release matched the measurements to an average of 94%. Thus, the two-dimensional heat transfer model increases the versatility of the ELESTRES code to better simulate fuels at normal as well as at extended burnups. (author)

  14. 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)

  15. ThO{sub 2}-UO{sub 2} annular pins for high burnup fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caner, Marc; Dugan, Edward T

    2000-06-01

    The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium-uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k{sub {infinity}} from two-dimensional-symmetry 'snapshots'.

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

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

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

  17. Using Flow Electrodes in Multiple Reactors in Series for Continuous Energy Generation from Capacitive Mixing

    KAUST Repository

    Hatzell, Marta C.

    2014-12-09

    Efficient conversion of “mixing energy” to electricity through capacitive mixing (CapMix) has been limited by low energy recoveries, low power densities, and noncontinuous energy production resulting from intermittent charging and discharging cycles. We show here that a CapMix system based on a four-reactor process with flow electrodes can generate constant and continuous energy, providing a more flexible platform for harvesting mixing energy. The power densities were dependent on the flow-electrode carbon loading, with 5.8 ± 0.2 mW m–2 continuously produced in the charging reactor and 3.3 ± 0.4 mW m–2 produced in the discharging reactor (9.2 ± 0.6 mW m–2 for the whole system) when the flow-electrode carbon loading was 15%. Additionally, when the flow-electrode electrolyte ion concentration increased from 10 to 20 g L–1, the total power density of the whole system (charging and discharging) increased to 50.9 ± 2.5 mW m–2.

  18. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

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

    Science.gov (United States)

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

    1997-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  1. Effects of continuous positive airway pressure on energy balance regulation: a systematic review

    Science.gov (United States)

    Shechter, Ari

    2016-01-01

    Obesity is both a cause and a possible consequence of obstructive sleep apnoea (OSA), as OSA seems to affect parameters involved in energy balance regulation, including food intake, hormonal regulation of hunger/satiety, energy metabolism and physical activity. It is known that weight loss improves OSA, yet it remains unclear why continuous positive airway pressure (CPAP) often results in weight gain. The goal of this systematic review is to explore if and how CPAP affects the behaviour and/or metabolism involved in regulating energy balance. CPAP appears to correct for a hormonal profile characterised by abnormally high leptin and ghrelin levels in OSA, by reducing the circulating levels of each. This is expected to reduce excess food intake. However, reliable measures of food intake are lacking, and not yet sufficient to make conclusions. Although studies are limited and inconsistent, CPAP may alter energy metabolism, with reports of reductions in resting metabolic rate or sleeping metabolic rate. CPAP appears to not have an appreciable effect on altering physical activity levels. More work is needed to characterise how CPAP affects energy balance regulation. It is clear that promoting CPAP in conjunction with other weight loss approaches should be used to encourage optimal outcomes in OSA patients. PMID:27824596

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

    Energy Technology Data Exchange (ETDEWEB)

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in

  3. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    various locations and at varying degrees during BWR operation based on the core loading pattern. When present during depletion, control blades harden the neutron spectrum locally because they displace the moderator and absorb thermal neutrons. The investigation of the effect of control blades on post operational cask reactivity is documented herein, as is the effect of multiple (continuous and intermittent) exposure periods with control blades inserted. The coupled effects of control blade presence on power density, void profile, or burnup profile will be addressed in future work.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

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

  5. Recovery of distortion product otoacoustic emissions (DPOAE) after impulse vs. continuous equal-energy exposures

    DEFF Research Database (Denmark)

    de Toro, Miguel Angel Aranda; Ordoñez, Rodrigo Pizarro; Reuter, Karen;

    2008-01-01

    -damage risk-criteria suffer from lack of empirical data needed to quantify impulse noise exposures and assess potential damage. In this experiment human subjects are exposed to binaural recordings of noises from industrial environments. Stimuli consist of impulse noise, continuous noise, and combinations......The correct assessment of impulse noise from occupational environments for hearing-conservation purposes is still a controversial issue. Currently, no universally accepted standard defines impulse noise accurately nor does a standard method exist to measure impulses. Moreover, current impulse...... of impulse and continuous noise. Noise exposures are normalized to have the same energy (LAeq,8h= 80dB). The effects in the hearing of the subjects are monitored by measuring the recovery of the distortion product otoacoustic emissions (DPOAE) with high-time resolution. The results can be used to investigate...

  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. 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.)

  8. Development of continuous-energy eigenvalue sensitivity coefficient calculation methods in the shift Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, C.; Martin, W. [Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States); Rearden, B.; Williams, M. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Div., Bldg. 5700, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)

    2012-07-01

    Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)

  9. Calculating kinetics parameters and reactivity changes with continuous-energy Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Kiedrowski, Brian C [Los Alamos National Laboratory; Brown, Forrest B [Los Alamos National Laboratory; Wilson, Paul [UNIV. WISCONSIN

    2009-01-01

    The iterated fission probability interpretation of the adjoint flux forms the basis for a method to perform adjoint weighting of tally scores in continuous-energy Monte Carlo k-eigenvalue calculations. Applying this approach, adjoint-weighted tallies are developed for two applications: calculating point reactor kinetics parameters and estimating changes in reactivity from perturbations. Calculations are performed in the widely-used production code, MCNP, and the results of both applications are compared with discrete ordinates calculations, experimental measurements, and other Monte Carlo calculations.

  10. Minimum Energy Control of 2D Positive Continuous-Discrete Linear Systems

    Directory of Open Access Journals (Sweden)

    Kaczorek Tadeusz

    2014-09-01

    Full Text Available The minimum energy control problem for the 2D positive continuous-discrete linear systems is formulated and solved. Necessary and sufficient conditions for the reachability at the point of the systems are given. Sufficient conditions for the existence of solution to the problem are established. It is shown that if the system is reachable then there exists an optimal input that steers the state from zero boundary conditions to given final state and minimizing the performance index for only one step (q = 1. A procedure for solving of the problem is proposed and illustrated by a numerical example.

  11. Continual Energy Management System of Proton Exchange Membrane Fuel Cell Hybrid Power Electric Vehicles

    Directory of Open Access Journals (Sweden)

    Ren Yuan

    2016-01-01

    Full Text Available Current research status in energy management of Proton Exchange Membrane (PEM fuel cell hybrid power electric vehicles are first described in this paper, and then build the PEMFC/ lithium-ion battery/ ultra-capacitor hybrid system model. The paper analysis the key factors of the continuous power available in PEM fuel cell hybrid power electric vehicle and hybrid power system working status under different driving modes. In the end this paper gives the working flow chart of the hybrid power system and concludes the three items of the system performance analysis.

  12. Nuclear Material Accountability Applications of a Continuous Energy and Direction Gamma Ray Detector

    Energy Technology Data Exchange (ETDEWEB)

    David Gerts; Robert Bean; Marc Paff

    2010-07-01

    The Idaho National Laboratory has recently developed a detector system based on the principle of a Wilson cloud chamber that gives the original energy and direction to a gamma ray source. This detector has the properties that the energy resolution is continuous and the direction to the source can be resolved to desired fidelity. Furthermore, the detector has low power requirements, is durable, operates in widely varying environments, and is relatively cheap to produce. This detector is expected, however, to require significant time to perform measurements. To mitigate the significant time for measurements, the detector is expected to scale to very large sizes with a linear increase in cost. For example, the proof of principle detector is approximately 30,000 cm3. This work describes the technical results that lead to these assertions. Finally, the applications of this detector are described in the context of nuclear material accountability.

  13. Syllable Segmentation of Farsi Continuous Speech using Wavelet Coefficients Thresholding and Fuzzy Smoothing of Energy Contour

    Directory of Open Access Journals (Sweden)

    Ghazaal Sheikhi

    2013-10-01

    Full Text Available Syllable, as a sub-word unit, nowadays plays an active role in the field of speech processing and recognition research according to its robust relation to human speech production and cognition. Automatic syllable boundaries detection is an important step forward in the areas of speech prosody, natural speech synthesis and speech recognition. In this paper, a novel method in automatic syllabification of Farsi continuous speech based on acoustic structure is proposed. Our previous studies, showed the proficiency of energy contour fuzzy smoothing method, compared with other prominent works in this area. This paper suggests that the conventional methodology-used in speech enhancement based on wavelet coefficient thresholding would improve syllable segmentation by decreasing insertion error. This process declines the energy in high energy consonants which are responsible for extra peaks in short term energy contour. Experimental results showed that utilizing proposed method along with fuzzy smoothing would diminish insertion error about 8% with no reasonable effect on other efficiency criteria. More than 94% of syllables are automatically segmented using presented technique with less than 50ms error.

  14. Continuity equations for bound electromagnetic field and the electromagnetic energy-momentum tensor

    Energy Technology Data Exchange (ETDEWEB)

    Kholmetskii, A L [Department of Physics, Belarusian State University, 4 Nezavisimosti Avenue, 220030 Minsk (Belarus); Missevitch, O V [Institute for Nuclear Problems, Belarusian State University, 11 Bobruiskaya Street, 220030 Minsk (Belarus); Yarman, T, E-mail: khol123@yahoo.com [Department of Engineering, Okan University, Akfirat, Istanbul, Turkey and Savronik, Eskisehir (Turkey)

    2011-05-01

    We analyze the application of the Poynting theorem to the bound (velocity-dependent) electromagnetic (EM) field and show that an often-used arbitrary elimination of the term of self-interaction in the product j{center_dot}E (where j is the current density and E the electric field) represents, in general, an illegitimate operation, which leads to incorrect physical consequences. We propose correct ways of eliminating the terms of self-interaction from the Poynting theorem to transform it into the form that is convenient for problems with bound EM field, which yield the continuity equations for the proper EM energy density, the interaction part of EM energy density and the total EM energy density of bound fields, respectively. These equations indicate the incompleteness of the common EM energy-momentum tensor, and in our analysis, we find a missed term in its structure, which makes its trace non-vanished. Some implications of these results are discussed, in particular, in view of the notion of EM mass of charged particles.

  15. Updated analytical solutions of continuity equation for electron beams precipitation - II. Mixed energy losses

    Science.gov (United States)

    Zharkova, V. V.; Dobranskis, R. R.

    2016-06-01

    In this paper we consider simultaneous analytical solutions of continuity equations for electron beam precipitation (a) in collisional losses and (b) in ohmic losses, or mixed energy losses (MEL) by applying the iterative method to calculate the resulting differential densities at given precipitation depth. The differential densities of precipitating electrons derived from the analytical solutions for MELs reveal increased flattening at energies below 10-30 keV compared to a pure collisional case. This flattening becomes stronger with an increasing precipitation depth turning into a positive slope at greater precipitation depths in the chromosphere resulting in a differential density distribution with maximum that shifts towards higher energies with increase in column depth, while the differential densities combining precipitating and returning electrons are higher at lower energies than those for a pure collisional case. The resulting hard X-ray (HXR) emission produced by the beams with different initial energy fluxes and spectral indices is calculated using the MEL approach for different ratios between the differential densities of precipitating and returning electrons. The number of returning electrons can be even further enhanced by a magnetic mirroring, not considered in the present model, while dominating at lower atmospheric depths where the magnetic convergence and magnitude are the highest. The proposed MEL approach provides an opportunity to account simultaneously for both collisional and ohmic losses in flaring events, which can be used for a quick spectral fitting of HXR spectra and evaluation of a fraction of returning electrons versus precipitating ones. The semi-analytical MEL approach is used for spectral fitting to Reuven High Energy Solar Spectroscopic Imager observations of nine C, M and X class flares revealing a close fit to the observations and good resemblance to numerical FP solutions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.O.

    1997-01-01

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

  18. Enhancements in Continuous-Energy Monte Carlo Capabilities for SCALE 6.2

    Energy Technology Data Exchange (ETDEWEB)

    Rearden, Bradley T [ORNL; Petrie Jr, Lester M [ORNL; Peplow, Douglas E. [ORNL; Bekar, Kursat B [ORNL; Wiarda, Dorothea [ORNL; Celik, Cihangir [ORNL; Perfetti, Christopher M [ORNL; Dunn, Michael E [ORNL

    2014-01-01

    SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, industry, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a plug-and-play framework that includes three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 provides several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, sensitivity and uncertainty analysis, and improved fidelity in nuclear data libraries. A brief overview of SCALE capabilities is provided with emphasis on new features for SCALE 6.2.

  19. Medium-sized grazing incidence high-energy X-ray telescopes employing continuously graded multilayers

    DEFF Research Database (Denmark)

    Joensen, K. D.; Christensen, Finn Erland; Schnopper, H. W.;

    1993-01-01

    The authors present a concept of continuously graded multilayer structures for medium-sized X-ray telescopes which is based on several material combinations. They show that the theoretical reflectivity characteristics of these structures make them very advantageous when applied to high energy X......-ray grazing incidence telescopes. They consider the performance of continuously graded Ni/C multilayers in a multi-focus, Kirkpatrick-Baez, geometry and show a significant improvement when compared to standard coatings of gold. For a total length of 3.3 m, a total aperture of 48 cm by 48 cm and 64 foci......, an effective area of 250 cm2 at 60 keV and a FWHM field of view of 6' is obtained. It is shown that a modular array of conical telescopes (conical approximation to a Wolter-I geometry), with the same length and aperture provides similar effective areas. Energy-dispersive X-ray reflectivity data (15-70 ke...

  20. Continuing California drought: an assessment of its effect on past and future energy production

    Energy Technology Data Exchange (ETDEWEB)

    1977-10-01

    The two successive dry years have severely impacted California's electrical utilities, and have caused them to make some major operating changes in order to provide adequate supplies of energy to meet their customer's demands. The following information describes the impact that two years of drought has had on the generating capabilities of the electric utilities in California, the contingency measures that have been taken, the minor events which have occasionally resulted in unusually low electrical reserve capacity margins, and what can be expected during 1978, should the drought continue. The staff of the Energy Commission conclude that by treating the entire state as a whole rather than by individual service areas, doing everything possible to maximize generation, and facilitating the transfer of power to areas of greatest need, that the State of California can withstand the threat of rolling blackouts, under the adverse conditions, without undue loss of income or substantial threat to the public health or safety. A continued drought will severely test the resources of the state's electrical utilities, and unusually high forced outage levels could conceivable create situations where occasional preplanned outages could result. However, increased conservation practices, proper advanced planning, complete cooperation between all utilities, and the maximum utilization of all generation resources, as recommended, should create a manageable situation in 1978.

  1. Energy-Efficient Integration of Continuous Context Sensing and Prediction into Smartwatches

    Directory of Open Access Journals (Sweden)

    Reza Rawassizadeh

    2015-09-01

    Full Text Available As the availability and use of wearables increases, they are becoming a promising platform for context sensing and context analysis. Smartwatches are a particularly interesting platform for this purpose, as they offer salient advantages, such as their proximity to the human body. However, they also have limitations associated with their small form factor, such as processing power and battery life, which makes it difficult to simply transfer smartphone-based context sensing and prediction models to smartwatches. In this paper, we introduce an energy-efficient, generic, integrated framework for continuous context sensing and prediction on smartwatches. Our work extends previous approaches for context sensing and prediction on wrist-mounted wearables that perform predictive analytics outside the device. We offer a generic sensing module and a novel energy-efficient, on-device prediction module that is based on a semantic abstraction approach to convert sensor data into meaningful information objects, similar to human perception of a behavior. Through six evaluations, we analyze the energy efficiency of our framework modules, identify the optimal file structure for data access and demonstrate an increase in accuracy of prediction through our semantic abstraction method. The proposed framework is hardware independent and can serve as a reference model for implementing context sensing and prediction on small wearable devices beyond smartwatches, such as body-mounted cameras.

  2. Energy-Efficient Integration of Continuous Context Sensing and Prediction into Smartwatches.

    Science.gov (United States)

    Rawassizadeh, Reza; Tomitsch, Martin; Nourizadeh, Manouchehr; Momeni, Elaheh; Peery, Aaron; Ulanova, Liudmila; Pazzani, Michael

    2015-09-08

    As the availability and use of wearables increases, they are becoming a promising platform for context sensing and context analysis. Smartwatches are a particularly interesting platform for this purpose, as they offer salient advantages, such as their proximity to the human body. However, they also have limitations associated with their small form factor, such as processing power and battery life, which makes it difficult to simply transfer smartphone-based context sensing and prediction models to smartwatches. In this paper, we introduce an energy-efficient, generic, integrated framework for continuous context sensing and prediction on smartwatches. Our work extends previous approaches for context sensing and prediction on wrist-mounted wearables that perform predictive analytics outside the device. We offer a generic sensing module and a novel energy-efficient, on-device prediction module that is based on a semantic abstraction approach to convert sensor data into meaningful information objects, similar to human perception of a behavior. Through six evaluations, we analyze the energy efficiency of our framework modules, identify the optimal file structure for data access and demonstrate an increase in accuracy of prediction through our semantic abstraction method. The proposed framework is hardware independent and can serve as a reference model for implementing context sensing and prediction on small wearable devices beyond smartwatches, such as body-mounted cameras.

  3. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

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

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

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

  5. Sculpting proteins interactively: continual energy minimization embedded in a graphical modeling system.

    Science.gov (United States)

    Surles, M C; Richardson, J S; Richardson, D C; Brooks, F P

    1994-02-01

    We describe a new paradigm for modeling proteins in interactive computer graphics systems--continual maintenance of a physically valid representation, combined with direct user control and visualization. This is achieved by a fast algorithm for energy minimization, capable of real-time performance on all atoms of a small protein, plus graphically specified user tugs. The modeling system, called Sculpt, rigidly constrains bond lengths, bond angles, and planar groups (similar to existing interactive modeling programs), while it applies elastic restraints to minimize the potential energy due to torsions, hydrogen bonds, and van der Waals and electrostatic interactions (similar to existing batch minimization programs), and user-specified springs. The graphical interface can show bad and/or favorable contacts, and individual energy terms can be turned on or off to determine their effects and interactions. Sculpt finds a local minimum of the total energy that satisfies all the constraints using an augmented Lagrange-multiplier method; calculation time increases only linearly with the number of atoms because the matrix of constraint gradients is sparse and banded. On a 100-MHz MIPS R4000 processor (Silicon Graphics Indigo), Sculpt achieves 11 updates per second on a 20-residue fragment and 2 updates per second on an 80-residue protein, using all atoms except non-H-bonding hydrogens, and without electrostatic interactions. Applications of Sculpt are described: to reverse the direction of bundle packing in a designed 4-helix bundle protein, to fold up a 2-stranded beta-ribbon into an approximate beta-barrel, and to design the sequence and conformation of a 30-residue peptide that mimics one partner of a protein subunit interaction. Computer models that are both interactive and physically realistic (within the limitations of a given force field) have 2 significant advantages: (1) they make feasible the modeling of very large changes (such as needed for de novo design), and

  6. Derrida's Generalized Random Energy models; 4, Continuous state branching and coalescents

    CERN Document Server

    Bovier, A

    2003-01-01

    In this paper we conclude our analysis of Derrida's Generalized Random Energy Models (GREM) by identifying the thermodynamic limit with a one-parameter family of probability measures related to a continuous state branching process introduced by Neveu. Using a construction introduced by Bertoin and Le Gall in terms of a coherent family of subordinators related to Neveu's branching process, we show how the Gibbs geometry of the limiting Gibbs measure is given in terms of the genealogy of this process via a deterministic time-change. This construction is fully universal in that all different models (characterized by the covariance of the underlying Gaussian process) differ only through that time change, which in turn is expressed in terms of Parisi's overlap distribution. The proof uses strongly the Ghirlanda-Guerra identities that impose the structure of Neveu's process as the only possible asymptotic random mechanism.

  7. Proposal of experimental device for the continuous accumulation of primary energy in natural gas hydrates

    Science.gov (United States)

    Siažik, Ján; Malcho, Milan; Lenhard, Richard

    2016-11-01

    Hydrates of the natural gas in the lithosphere are a very important potential source of energy that will be probably used in the coming decades. It seems as promising accumulation of the standard gas to form hydrates synthetically, stored, and disengage him when is peak demand. Storage of natural gas or biomethane in hydrates is advantageous not only in terms of storage capacity, but also from the aspect of safety storage hydrates. The gas stored in such form may occurs at relatively high temperatures and low pressures in comparison to other Technologies of gas- storage. In one cubic meter of hydrate can be stored up to 150 m3 of natural gas, depending on the conditions of thermobaric hydrate generation. This article discusses the design of the facility for the continuous generation of hydrates of natural gas measurement methodology and optimal conditions for their generation.

  8. A knowledge continuity management program for the energy, infrastructure and knowledge systems center, Sandia National Laboratories.

    Energy Technology Data Exchange (ETDEWEB)

    Menicucci, David F.

    2006-07-01

    A growing recognition exists in companies worldwide that, when employees leave, they take with them valuable knowledge that is difficult and expensive to recreate. The concern is now particularly acute as the large ''baby boomer'' generation is reaching retirement age. A new field of science, Knowledge Continuity Management (KCM), is designed to capture and catalog the acquired knowledge and wisdom from experience of these employees before they leave. The KCM concept is in the final stages of being adopted by the Energy, Infrastructure, and Knowledge Systems Center and a program is being applied that should produce significant annual cost savings. This report discusses how the Center can use KCM to mitigate knowledge loss from employee departures, including a concise description of a proposed plan tailored to the Center's specific needs and resources.

  9. A correlated study between effective total macroscopic cross sections and effective energies for neutron beams with continuous spectra

    CERN Document Server

    Kobayashi, H

    1999-01-01

    Two practically useful quantities have been introduced to characterize a continuous-energy-spectrum neutron beam and to describe transmission phenomena of the beam in the field of quantitative neutron radiography. These quantities are the effective energy instead of a peak energy or a mean energy of the spectrum and an effective total macroscopic (ETM) cross section instead of a total macroscopic (TM) cross section as defined for a monochromatic energy. Four neutron beams have been used to measure ETM cross sections at effective energies of 29.8, 17.2, 9.8 meV, and at the In resonance energy of 1.46 eV. Results are studied as a function of estimated effective energy, where the effective energy was estimated by a beam quality indicator (BQI) which has been proposed recently. Validity of ETM cross sections as a function of the effective energy is discussed and correlated with recent nuclear data.

  10. Continuous distributed phase-plate advances for high-energy laser systems

    Science.gov (United States)

    Marozas, J. A.; Collins, T. J. B.; Zuegel, J. D.; McKenty, P. W.; Cao, D.; Fochs, S.; Radha, P. B.

    2016-05-01

    The distributed phase plate (DPP) design code Zhizhoo’ has been used to design full- aperture, continuous near-field transmission optics for a wide variety of high-fidelity focal-spot shapes for high-energy laser systems: OMEGA EP, Dynamic Compression Sector (DCS), and the National Ignition Facility (NIF). The envelope shape, or profile, of the focal spot affects the hydrodynamics of directly driven targets in these laser systems. Controlling the envelope shape to a high degree of fidelity impacts the quality of the ablatively driven implosions. The code Zhizhoo’ not only produces DPP's with great control of the envelope shape, but also spectral and gradient control as well as robustness from near-field phase aberrations. The focal-spot shapes can take on almost any profile from symmetric to irregular patterns and with high fidelity relative to the objective function over many decades of intensity. The control over the near-field phase spectrum and phase gradients offer greater manufacturability of the full- aperture continuous surface-relief pattern. The flexibility and speed of the DPP design code Zhizhoo’ will be demonstrated by showing the wide variety of successful designs that have been made and those that are in progress.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  12. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    Science.gov (United States)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  13. 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

  14. 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.

  15. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  16. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

    2016-11-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  17. Laboratory investigations on continuous bio-methanization of energy crops as mono-substrate without supplementation

    Energy Technology Data Exchange (ETDEWEB)

    Demirel, Burak [Hamburg University of Applied Sciences (HAW Hamburg), Lifetec Process Engineering, Lohbruegger Kirchstrasse 65, 21033 Hamburg-Bergedorf (Germany); Bogazici University, Institute of Environmental Sciences, Bebek, 34342 Istanbul (Turkey)

    2009-06-15

    Continuous bio-methanization of an energy crop, namely the beet silage, was investigated in this laboratory-scale work as mono-substrate, using a mesophilic biogas digester controlled by a fuzzy logic control (FLC) technique and without using any supplementing or buffering agent, despite the low pH of the substrate around 3.80. The temperature, pH, redox potential (ORP), daily biogas production and composition of digester biogas were continuously measured online. During the operation, the hydraulic retention time (HRT) varied between 24.8 and 9 days, as the organic loading rate (OLR) ranged from 2.6 to 4.7 g L{sup -1} d{sup -1}. The average pH, specific gas production rate (spec. GPR) and volumetric gas production rate (vol. GPR) were determined to be 7.12, 0.31 L g VS{sup -1} d{sup -1} and 1.084 L L{sup -1} d{sup -1}, respectively. The average methane (CH{sub 4}) content of digester biogas was about 56%. The FLC technique, which was developed at HAW Hamburg for anaerobic conversion of acidic energy crops to methane, determined the daily feeding volume ({proportional_to} OLR/HRT) for the biogas digester, depending on the feedback from online pH and methane measurements, and on the calculation of the spec. GPR. The spec. GPR was calculated by the corrected daily biogas production. Through online monitoring of pH, biogas production rate and composition, and by use of the FLC technique, the acidic beet silage could continuously be converted to biogas, without using manure or any other kind of buffering or supplementing agent(s). The lab-scale anaerobic biogas digester performed stable and safe, without encountering any problems of instability, as indicated by an adequate amount of buffering capacity, a VFA content below 0.5 g L{sup -1} and a neutral pH range throughout the study. (author)

  18. Process monitoring and modeling of a continuous pharmaceutical from powder to tablet process Line using a mass & energy balance

    DEFF Research Database (Denmark)

    Beer, Thomas De; Mortier, Séverine Thérèse F.C.; Gernaey, Krist

    2014-01-01

    The intention to shift from batch to continuous production processes within the pharmaceutical industry enhances the need to monitor and control the process in-line and real-time to continuously guarantee the end-product quality. Mass and energy balances have been successfully applied to a drying...

  19. Studi Eksperimen Analisa Performa Compact Heat Exchanger Circular Tubes Continuous Plate Fin Untuk Pemanfaatan Waste Energy

    Directory of Open Access Journals (Sweden)

    Rachmadi Gewa Saputra

    2014-03-01

    Full Text Available Harga minyak dunia cenderung mengalami peningkatan dalam beberapa tahun terakhir sehingga manusia berfikir untuk memanfaatkan setiap penggunaan minyak bumi. Dengan berkembangnya teknologi saat ini waste energy yang berupa gas hasil pembakaran pada engine dapat dimanfaatkan menjadi bentuk energi lain menggunakan heat recovery system. Pada tugas akhir ini dilakukan desain sebuah heat exchanger tipe circular tubes continuous plate fin dengan susunan tube aligned yang digunakan untuk menyerap waste energy yang berupa exhaust gas. Untuk mendapatkan dimensi desain yang sesuai digunakan metode ΔTLMTD. Metode ini digunakan untuk menentukan nilai dari overall heat transfer  desain dari heat exchanger, kemudian dilakukan perhitungan untuk nilai overall heat transfer hitung. Setelah didapatkan nilai dari overall heat transfer secara desain dan hitung maka dilakukan iterasi untuk mendapatkan dimensi heat exchanger yang memiliki nilai error paling kecil antara nilai overall heat transfer desain dan hitung. Untuk pengujian performa dari heat exchanger yang telah didesain maka dilakukan variasi kacepatan exhaust gas yang melewati heat exchanger, yaitu 0.4 m/s, 0.3 m/s, dan 0.2 m/s. Exhaust gas yang digunakan memiliki temperatur 280oC. Pada tugas akhir ini didapatkan desain compact heat exchanger dengan dimensi panjang 0.38 m, lebar 0.45 m, dan tebal 0.04m. Setelah dilakukan pengujian dengan memvariasikan kecepatan dari exhaust gas yang melewati heat exchanger maka didapatkan bahwa nilai dari qaktual dari heat exchanger mengalami kenaikan dengan bertambahnya reynolds number akibat bertambahnya kecepatan exhaust gas, kemudian nilai dari effectiveness akan mengalami penurunan untuk setiap kenaikan dari reynold number exhaust gas. Selain itu nilai dari NTU heat exchanger juga mengalami penurunan dengan bertambahnya reynold number exhaust gas. Untuk nilai overall heat transfer dari heat exchanger yang didesain akan mengalami kenaikan akibat bertambahnya nilai

  20. Composites (CFCCs) for low cost energy and cleaner environment. Continuous fiber ceramic composites program

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    For many industrial applications, materials are desired which combine light weight, high temperature strength, and stability in corrosive environments. Among competing materials, ceramics are noteworthy candidates for such applications. The use of ceramics is often constrained, however, by brittleness; i.e., low toughness. Ceramic composites are being developed to overcome this limitation. With recent advances in ceramic fiber technology, it is possible to design a composite material based on continuous ceramic fibers embedded in a ceramic matrix. The use of ceramic composites in industrial applications will result in reduced fuel consumption, but will also prevent airborne pollution (principally NO, SO{sub x}, CO{sub 2}, and particulates), and economically benefit the end user through energy and environmental savings and increased competitiveness. Industry will also benefit through increased productivity and consumers will benefit through lower energy and environmental costs and a cleaner environment. The development and use of CFCCs could become an important factor in the international competitiveness of U.S. industry. CFCCs will be a critical enabling material in the design and engineering of advanced components, systems, and processes. If CFCC technology is developed outside the United States, domestic users of these materials may be forced to rely on foreign suppliers of the products fabricated from CFCCs, as well as the materials themselves. Foreign countries, including Japan and France, have embarked on government-supported CFCC development efforts. With the market for CFCC products expected to be a $10 billion dollar market by 2010, CFCC development will be important for the competitiveness of U.S. industry and for retaining and creating jobs for U.S. citizens. This document summarizes the potential energy, environmental, and economic benefits that CFCCs will have for the U.S. economy and particularly for the industrial sector.

  1. Implementation of the probability table method in a continuous-energy Monte Carlo code system

    Energy Technology Data Exchange (ETDEWEB)

    Sutton, T.M.; Brown, F.B. [Lockheed Martin Corp., Schenectady, NY (United States)

    1998-10-01

    RACER is a particle-transport Monte Carlo code that utilizes a continuous-energy treatment for neutrons and neutron cross section data. Until recently, neutron cross sections in the unresolved resonance range (URR) have been treated in RACER using smooth, dilute-average representations. This paper describes how RACER has been modified to use probability tables to treat cross sections in the URR, and the computer codes that have been developed to compute the tables from the unresolved resonance parameters contained in ENDF/B data files. A companion paper presents results of Monte Carlo calculations that demonstrate the effect of the use of probability tables versus the use of dilute-average cross sections for the URR. The next section provides a brief review of the probability table method as implemented in the RACER system. The production of the probability tables for use by RACER takes place in two steps. The first step is the generation of probability tables from the nuclear parameters contained in the ENDF/B data files. This step, and the code written to perform it, are described in Section 3. The tables produced are at energy points determined by the ENDF/B parameters and/or accuracy considerations. The tables actually used in the RACER calculations are obtained in the second step from those produced in the first. These tables are generated at energy points specific to the RACER calculation. Section 4 describes this step and the code written to implement it, as well as modifications made to RACER to enable it to use the tables. Finally, some results and conclusions are presented in Section 5.

  2. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  3. Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.

    Science.gov (United States)

    Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K

    2013-10-01

    Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup.

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

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

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

  5. Bio-methanization of energy crops through mono-digestion for continuous production of renewable biogas

    Energy Technology Data Exchange (ETDEWEB)

    Demirel, Burak [Lifetec Process Engineering, Hamburg University of Applied Sciences, Lohbruegger Kirchstrasse 65, 21033 Hamburg-Bergedorf (Germany); Bogazici University, Institute of Environmental Sciences, Bebek 34342, Istanbul (Turkey); Scherer, Paul [Lifetec Process Engineering, Hamburg University of Applied Sciences, Lohbruegger Kirchstrasse 65, 21033 Hamburg-Bergedorf (Germany)

    2009-12-15

    The aim of this laboratory-scale study was to investigate the long-term anaerobic fermentation of an extremely sour substrate, an energy crop, for continuous production of methane (CH{sub 4}) as a source of renewable energy. The sugar beet silage was used as the mono-substrate, which had a low pH of around 3.3-3.4, without the addition of manure. The mesophilic biogas digester was operated in a hydraulic retention time (HRT) range between 15 and 9.5 days, and an organic loading rate (OLR) range of between 6.33 and 10 g VS l{sup -1} d{sup -1}. The highest specific gas production rate (spec. GPR) and CH{sub 4} content were 0.67 l g VS{sup -1} d{sup -1} and 74%, respectively, obtained at an HRT of 9.5 days and OLR of 6.35 g VS l{sup -1} d{sup -1}. The digester worked within the neutral pH range as well. Since this substrate lacked the availability of macro and micro nutrients, and the buffering capacity as well, external supplementation was definitely required to provide a stable and efficient operation, as provided using NH{sub 4}Cl and KHCO{sub 3} in this case. The findings of this ongoing long-term fermentation of an extremely acidic biomass substrate without manure addition have reflected crucial information about how to appropriately maintain the operational and particularly the environmental parameters in an agricultural biogas plant. (author)

  6. SCALE Continuous-Energy Monte Carlo Depletion with Parallel KENO in TRITON

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, Sedat [ORNL; Bekar, Kursat B [ORNL; Wiarda, Dorothea [ORNL

    2012-01-01

    The TRITON sequence of the SCALE code system is a powerful and robust tool for performing multigroup (MG) reactor physics analysis using either the 2-D deterministic solver NEWT or the 3-D Monte Carlo transport code KENO. However, as with all MG codes, the accuracy of the results depends on the accuracy of the MG cross sections that are generated and/or used. While SCALE resonance self-shielding modules provide rigorous resonance self-shielding, they are based on 1-D models and therefore 2-D or 3-D effects such as heterogeneity of the lattice structures may render final MG cross sections inaccurate. Another potential drawback to MG Monte Carlo depletion is the need to perform resonance self-shielding calculations at each depletion step for each fuel segment that is being depleted. The CPU time and memory required for self-shielding calculations can often eclipse the resources needed for the Monte Carlo transport. This summary presents the results of the new continuous-energy (CE) calculation mode in TRITON. With the new capability, accurate reactor physics analyses can be performed for all types of systems using the SCALE Monte Carlo code KENO as the CE transport solver. In addition, transport calculations can be performed in parallel mode on multiple processors.

  7. Continuous energy Monte Carlo calculations for randomly distributed spherical fuels based on statistical geometry model

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Isao [Osaka Univ., Suita (Japan); Mori, Takamasa; Nakagawa, Masayuki; Itakura, Hirofumi

    1996-03-01

    The method to calculate neutronics parameters of a core composed of randomly distributed spherical fuels has been developed based on a statistical geometry model with a continuous energy Monte Carlo method. This method was implemented in a general purpose Monte Carlo code MCNP, and a new code MCNP-CFP had been developed. This paper describes the model and method how to use it and the validation results. In the Monte Carlo calculation, the location of a spherical fuel is sampled probabilistically along the particle flight path from the spatial probability distribution of spherical fuels, called nearest neighbor distribution (NND). This sampling method was validated through the following two comparisons: (1) Calculations of inventory of coated fuel particles (CFPs) in a fuel compact by both track length estimator and direct evaluation method, and (2) Criticality calculations for ordered packed geometries. This method was also confined by applying to an analysis of the critical assembly experiment at VHTRC. The method established in the present study is quite unique so as to a probabilistic model of the geometry with a great number of spherical fuels distributed randomly. Realizing the speed-up by vector or parallel computations in future, it is expected to be widely used in calculation of a nuclear reactor core, especially HTGR cores. (author).

  8. Simulation of Watts Bar Unit 1 Initial Startup Tests with Continuous Energy Monte Carlo Methods

    Energy Technology Data Exchange (ETDEWEB)

    Godfrey, Andrew T [ORNL; Gehin, Jess C [ORNL; Bekar, Kursat B [ORNL; Celik, Cihangir [ORNL

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors* is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications. One component of the testing and validation plan for VERA is comparison of neutronics results to a set of continuous energy Monte Carlo solutions for a range of pressurized water reactor geometries using the SCALE component KENO-VI developed by Oak Ridge National Laboratory. Recent improvements in data, methods, and parallelism have enabled KENO, previously utilized predominately as a criticality safety code, to demonstrate excellent capability and performance for reactor physics applications. The highly detailed and rigorous KENO solutions provide a reliable nu-meric reference for VERAneutronics and also demonstrate the most accurate predictions achievable by modeling and simulations tools for comparison to operating plant data. This paper demonstrates the performance of KENO-VI for the Watts Bar Unit 1 Cycle 1 zero power physics tests, including reactor criticality, control rod worths, and isothermal temperature coefficients.

  9. Continuous operation of an ultra-low-power microcontroller using glucose as the sole energy source.

    Science.gov (United States)

    Lee, Inyoung; Sode, Takashi; Loew, Noya; Tsugawa, Wakako; Lowe, Christopher Robin; Sode, Koji

    2017-07-15

    An ultimate goal for those engaged in research to develop implantable medical devices is to develop mechatronic implantable artificial organs such as artificial pancreas. Such devices would comprise at least a sensor module, an actuator module, and a controller module. For the development of optimal mechatronic implantable artificial organs, these modules should be self-powered and autonomously operated. In this study, we aimed to develop a microcontroller using the BioCapacitor principle. A direct electron transfer type glucose dehydrogenase was immobilized onto mesoporous carbon, and then deposited on the surface of a miniaturized Au electrode (7mm(2)) to prepare a miniaturized enzyme anode. The enzyme fuel cell was connected with a 100 μF capacitor and a power boost converter as a charge pump. The voltage of the enzyme fuel cell was increased in a stepwise manner by the charge pump from 330mV to 3.1V, and the generated electricity was charged into a 100μF capacitor. The charge pump circuit was connected to an ultra-low-power microcontroller. Thus prepared BioCapacitor based circuit was able to operate an ultra-low-power microcontroller continuously, by running a program for 17h that turned on an LED every 60s. Our success in operating a microcontroller using glucose as the sole energy source indicated the probability of realizing implantable self-powered autonomously operated artificial organs, such as artificial pancreas. Copyright © 2016 Elsevier B.V. All rights reserved.

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

    CERN Document Server

    Ohki, S

    2003-01-01

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

  11. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  12. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  13. Reactivity effect of spent fuel depending on burn-up history

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takafumi [Nagoya Univ., Nagoya, Aichi (Japan); Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Mochizuki, Hiroki [The Japan Research Institute, Ltd., Tokyo (Japan)

    2001-06-01

    It is well known that a composition of spent fuel depends on various parameter changes throughout a burn-up period. In this study we aimed at the boron concentration and its change, the coolant temperature and its spatial distribution, the specific power, the operation mode, and the duration of inspection, because the effects due to these parameters have not been analyzed in detail. The composition changes of spent fuel were calculated by using the burn-up code SWAT, when the parameters mentioned above varied in the range of actual variations. Moreover, to estimate the reactivity effect caused by the composition changes, the criticality calculations for an infinite array of spent fuel were carried out with computer codes SRAC95 or MVP. In this report the reactivity effects were arranged from the viewpoint of what parameters gave more positive reactivity effect. The results obtained through this study are useful to choose the burn-up calculation model when we take account of the burn-up credit in the spent fuel management. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-01

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

  16. Comparison between SERPENT and MONTEBURNS codes applied to burnup calculations of a GFR-like configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chersola, Davide [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Marotta, Riccardo [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Mazzini, Guido [Centrum výzkumu Řež (Research Centre Rez), Husinec-Rez, cp. 130, 25068 Rez (Czech Republic)

    2014-07-01

    Highlights: • MC codes are widely adopted to analyze nuclear facilities, including GEN-IV reactors. • Burnup calculations are an efficient tool to test neutronic Monte Carlo codes. • In this comparison the used codes show some differences but a good agreement exists. - Abstract: This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MW{sub th} reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

  17. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-15

    Highlights: • Thermal properties of irradiated U–Mo alloy monolithic fuel samples were measured. • Density, thermal diffusivity, and thermal conductivity are influenced by increasing burnup. • U–Mo chemistry and specific heat capacity was not as sensitive to increasing burnup. • Thermal conductivity decreased approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} at 200 °C. • An empirical model developed previously agrees well with the experimental measurements. - Abstract: A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U–Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U–Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U–Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U–Mo alloy decreased approximately 30% for a fission density of 3.30 × 10{sup 21} fissions cm{sup −3} and approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  18. Oxygen potential measurements in high burnup LWR U0 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

  19. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  20. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Science.gov (United States)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

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

  2. Controlled passive actuation: concepts for energy efficient actuation using mechanical storage elements and continuously variable transmissions

    NARCIS (Netherlands)

    Dresscher, Douwe

    2016-01-01

    Walking robots consume more energy for locomotion than their wheeled and tracked counterparts. To achieve energy autonomous operation, a robot needs to run on energy that is harvested from its environment. In this light, it is meaningful to address reduction of energy consumption. The contribution

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  4. Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

    2008-04-15

    Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

  6. SCALE 6.2 Continuous-Energy TSUNAMI-3D Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Perfetti, Christopher M [ORNL; Rearden, Bradley T [ORNL

    2015-01-01

    The TSUNAMI (Tools for Sensitivity and UNcertainty Analysis Methodology Implementation) capabilities within the SCALE code system make use of sensitivity coefficients for an extensive number of criticality safety applications, such as quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different systems, quantifying computational biases, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved ease of use and fidelity and the desire to extend TSUNAMI analysis to advanced applications have motivated the development of a SCALE 6.2 module for calculating sensitivity coefficients using three-dimensional (3D) continuous-energy (CE) Monte Carlo methods: CE TSUNAMI-3D. This paper provides an overview of the theory, implementation, and capabilities of the CE TSUNAMI-3D sensitivity analysis methods. CE TSUNAMI contains two methods for calculating sensitivity coefficients in eigenvalue sensitivity applications: (1) the Iterated Fission Probability (IFP) method and (2) the Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance CHaracterization (CLUTCH) method. This work also presents the GEneralized Adjoint Response in Monte Carlo method (GEAR-MC), a first-of-its-kind approach for calculating adjoint-weighted, generalized response sensitivity coefficients—such as flux responses or reaction rate ratios—in CE Monte Carlo applications. The accuracy and efficiency of the CE TSUNAMI-3D eigenvalue sensitivity methods are assessed from a user perspective in a companion publication, and the accuracy and features of the CE TSUNAMI-3D GEAR-MC methods are detailed in this paper.

  7. Effects of continuous-wave, pulsed, and sinusoidal-amplitude-modulated microwaves on brain energy metabolism.

    Science.gov (United States)

    Sanders, A P; Joines, W T; Allis, J W

    1985-01-01

    A comparison of the effects of continuous-wave, sinusoidal-amplitude-modulated, and pulsed square-wave-modulated 591-MHz microwave exposures on brain energy metabolism was made in male Sprague-Dawley rats (175-225 g). Brain NADH fluorescence, adenosine triphosphate (ATP) concentration, and creatine phosphate (CP) concentration were determined as a function of modulation frequency. Brain temperatures of animals were maintained between -0.1 and -0.4 degrees C from the preexposure temperature when subjected to as much as 20 mW/cm2 (average power) CW, pulsed, or sinusoidal-amplitude modulated 591-MHz radiation for 5 min. Sinusoidal-amplitude-modulated exposures at 16-24 Hz showed a trend toward preferential modulation frequency response in inducing an increase in brain NADH fluorescence. The pulse-modulated and sinusoidal-amplitude-modulated (16 Hz) microwaves were not significantly different from CW exposures in inducing increased brain NADH fluorescence and decreased ATP and CP concentrations. When the pulse-modulation frequency was decreased from 500 to 250 pulses per second the average incident power density threshold for inducing an increase in brain NADH fluorescence increased by a factor of 4--ie, from about 0.45 to about 1.85 mW/cm2. Since brain temperature did not increase, the microwave-induced increase in brain NADH and decrease in ATP and CP concentrations was not due to hyperthermia. This suggests a direct interaction mechanism and is consistent with the hypothesis of microwave inhibition of mitochondrial electron transport chain function of ATP production.

  8. Ageing first passage time density in continuous time random walks and quenched energy landscapes

    Science.gov (United States)

    Krüsemann, Henning; Godec, Aljaž; Metzler, Ralf

    2015-07-01

    We study the first passage dynamics of an ageing stochastic process in the continuous time random walk (CTRW) framework. In such CTRW processes the test particle performs a random walk, in which successive steps are separated by random waiting times distributed in terms of the waiting time probability density function \\psi (t)≃ {t}-1-α (0≤slant α ≤slant 2). An ageing stochastic process is defined by the explicit dependence of its dynamic quantities on the ageing time ta, the time elapsed between its preparation and the start of the observation. Subdiffusive ageing CTRWs with 0\\lt α \\lt 1 describe systems such as charge carriers in amorphous semiconducters, tracer dispersion in geological and biological systems, or the dynamics of blinking quantum dots. We derive the exact forms of the first passage time density for an ageing subdiffusive CTRW in the semi-infinite, confined, and biased case, finding different scaling regimes for weakly, intermediately, and strongly aged systems: these regimes, with different scaling laws, are also found when the scaling exponent is in the range 1\\lt α \\lt 2, for sufficiently long ta. We compare our results with the ageing motion of a test particle in a quenched energy landscape. We test our theoretical results in the quenched landscape against simulations: only when the bias is strong enough, the correlations from returning to previously visited sites become insignificant and the results approach the ageing CTRW results. With small bias or without bias, the ageing effects disappear and a change in the exponent compared to the case of a completely annealed landscape can be found, reflecting the build-up of correlations in the quenched landscape.

  9. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    Science.gov (United States)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  10. 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.

  11. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Science.gov (United States)

    Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

    2013-10-01

    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.

  12. Burnup calculation by the method of first-flight collision probabilities using average chords prior to the first collision

    Science.gov (United States)

    Karpushkin, T. Yu.

    2012-12-01

    A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  14. Energy data visualisation requires additional approaches to continue to be relevant in a world with greater low-carbon generation.

    Directory of Open Access Journals (Sweden)

    I.A. Grant Wilson

    2016-08-01

    Full Text Available The hypothesis described in this article proposes that energy visualisation diagrams commonly used need additional changes to continue to be relevant in a world with greater low-carbon generation. The diagrams that display national energy data are influenced by the properties of the type of energy being displayed, which in most cases has historically meant fossil fuels, nuclear fuels or hydro. As many energy systems throughout the world increase their use of electricity from wind or solar based renewables, a more granular display of energy data in the time domain is required. This article also introduces the shared axes energy diagram that provides a simple and powerful way in which to compare the scale and seasonality of the demands and supplies of an energy system. This aims to complement rather than replace existing diagrams, and has an additional benefit of promoting a whole systems approach to energy systems, as differing energy vectors such as natural gas, transport fuels, and electricity can all be displayed together. This in particular, is useful to both policy makers and to industry, to build a visual foundation for a whole systems narrative, which provides a basis for discussion of the synergies and opportunities across and between different energy vectors and demands. The diagram’s ability to wrap a sense of scale around a whole energy system in a simple way is thought to explain its growing popularity.

  15. Estimating Energy- and Eco-Balances for Continuous Bio-Ethanol Production Using a Blenke Cascade System

    OpenAIRE

    2013-01-01

    Energy and environmental effects of wheat-based fuel, produced continuously by a Blenke cascade system, were assessed. Two scenarios: (1) no-co-products utilization scenario; and (2) co-products utilization scenario, were compared. A Life Cycle Assessment (LCA) model was used for analysis. The scope covered a cradle-to-gate inventory. The results from energy analysis showed, that wheat-based ethanol has a positive average net energy value (NEV), NEV = 3.35 MJ/kg ethanol with an average net en...

  16. Development of a Burnup Module DECBURN Based on the Krylov Subspace Method

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Kim, K. S.; Shim, H. J.; Song, J. S

    2008-05-15

    This report is to develop a burnup module DECBURN that is essential for the reactor analysis and the assembly homogenization codes to trace the fuel composition change during the core burnup. The developed burnup module solves the burnup equation by the matrix exponential method based on the Krylov Subspace method. The final solution of the matrix exponential is obtained by the matrix scaling and squaring method. To develop DECBURN module, this report includes the followings as: (1) Krylov Subspace Method for Burnup Equation, (2) Manufacturing of the DECBURN module, (3) Library Structure Setup and Library Manufacturing, (4) Examination of the DECBURN module, (5) Implementation to the DeCART code and Verification. DECBURN library includes the decay constants, one-group cross section and the fission yields. Examination of the DECBURN module is performed by manufacturing a driver program, and the results of the DECBURN module is compared with those of the ORIGEN program. Also, the implemented DECBURN module to the DeCART code is applied to the LWR depletion benchmark and a OPR-1000 pin cell problem, and the solutions are compared with the HELIOS code to verify the computational soundness and accuracy. In this process, the criticality calculation method and the predictor-corrector scheme are introduced to the DeCART code for a function of the homogenization code. The examination by a driver program shows that the DECBURN module produces exactly the same solution with the ORIGEN program. DeCART code that equips the DECBURN module produces a compatible solution to the other codes for the LWR depletion benchmark. Also the multiplication factors of the DeCART code for the OPR-1000 pin cell problem agree to the HELIOS code within 100 pcm over the whole burnup steps. The multiplication factors with the criticality calculation are also compatible with the HELIOS code. These results mean that the developed DECBURN module works soundly and produces an accurate solution

  17. Update of Continuous-Energy Data for Hydrogen and SiO2 Thermal Scattering

    Energy Technology Data Exchange (ETDEWEB)

    Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald Kent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-23

    The Nuclear Data Team has released updated continuous-energy neutron data files for: 1) hydrogen, and 2) S (α; β) (thermal scattering) on SiO2. A list of new ZAIDs and the data that is updated (Old ZAID) is given in Table 1. The old data are still accessible, but are not the default.

  18. An Overview of Two Years of Continuous Energy Optimization at the Velenje Coal Mine

    Directory of Open Access Journals (Sweden)

    Milan Medved

    2012-06-01

    Full Text Available The Velenje Coal Mine (VCM is one of the largest and the most modern underground coal mines in Europe. Although the coal mining industry produces coal as an energy source, it is also uses a lot of energy for its own operation and support processes. At this time of volatile energy prices and more and more strict environmental emission requirements, optimizing energy consumption plays an important role in good business performance. To track the consumption of electricity, district heating, drinking water and compressed air at the VCM a detailed energy monitoring methodology was developed and established in July 2010. The essential element of the presented monitoring system is a software application named “Central System for Regulation of Energy” (CSRE. The purpose of the CSRE is to control energy processes from a distance, take measures for economical and efficient use of energy, as well as to assist in maintenance. Such monitoring allows extensive comparisons between different energy sources consumption and enables correct measures to be taken to reduce the difference between the target and actual consumption of energy in VCM. With established real-time monitoring system, it is possible to look at mining processes and see where energy is being used inefficiently.

  19. Throughput Maximization for Sensor-Aided Cognitive Radio Networks with Continuous Energy Arrivals.

    Science.gov (United States)

    Nguyen, Thanh-Tung; Koo, Insoo

    2015-11-27

    We consider a Sensor-Aided Cognitive Radio Network (SACRN) in which sensors capable of harvesting energy are distributed throughout the network to support secondary transmitters for sensing licensed channels in order to improve both energy and spectral efficiency. Harvesting ambient energy is one of the most promising solutions to mitigate energy deficiency, prolong device lifetime, and partly reduce the battery size of devices. So far, many works related to SACRN have considered single secondary users capable of harvesting energy in whole slot as well as short-term throughput. In the paper, we consider two types of energy harvesting sensor nodes (EHSN): Type-I sensor nodes will harvest ambient energy in whole slot duration, whereas type-II sensor nodes will only harvest energy after carrying out spectrum sensing. In the paper, we also investigate long-term throughput in the scheduling window, and formulate the throughput maximization problem by considering energy-neutral operation conditions of type-I and -II sensors and the target detection probability. Through simulations, it is shown that the sensing energy consumption of all sensor nodes can be efficiently managed with the proposed scheme to achieve optimal long-term throughput in the window.

  20. Double unification of particles with fields and electricity with gravity in non-empty space of continuous complex energies

    Science.gov (United States)

    Bulyzhenkov, Igor E.

    2016-11-01

    Non-empty space reading of Maxwell equations as local energy identities explains why a Coulomb field is carried rigidly by electrons in experiments. The analytical solution of the Poisson equation defines the sharp radial shape of charged elementary densities which are proportional to continuous densities of electric self-energy. Both Coulomb field and radial charge densities are free from energy divergences. Non-empty space of electrically charged mass-energy can be described by complex analytical densities resulting in real values for volume mass integrals and in imaginary values for volume charge integrals. Imaginary electric charges in the Newton gravitational law comply with real Coulomb forces. Unification of forces through complex charges rids them of radiation self-acceleration. Strong gravitational fields repeal probe bodies that might explainthe accelerated expansion of the dense Metagalaxy. Outward and inward spherical waves form the standing wave process within the radial carrier of complex energy.

  1. High Energy and Power Co-­Continuous Electrodes Derived from Bijels

    OpenAIRE

    Witt, Jessica

    2015-01-01

    Next-generation energy storage materials aim to combine the best characteristics of batteries and supercapacitors, producing devices that concurrently deliver high energy and power densities. In order to achieve this goal, the kinetics of ion and electron transport within the electrodes must be enhanced while maintaining a large volume fraction of electrolytically active material for energy storage. In this regard, the idealized electrode microstructure has been envisioned as a three-dimensio...

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

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

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

  3. The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; John Maki

    2005-02-01

    The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

  4. On Energy Functions for String-Like Continuous Curves, Discrete Chains, and Space-Filling One Dimensional Structures

    CERN Document Server

    Hu, Shuangwei; Niemi, Antti J

    2012-01-01

    The theory of string-like continuous curves and discrete chains have numerous important physical applications. Here we develop a general geometrical approach, to systematically derive Hamiltonian energy functions for these objects. In the case of continuous curves, we demand that the energy function must be invariant under local frame rotations, and it should also transform covariantly under reparametrizations of the curve. This leads us to consider energy functions that are constructed from the conserved quantities in the hierarchy of the integrable nonlinear Schr\\"odinger equation (NLSE). We point out the existence of a Weyl transformation that we utilize to introduce a dual hierarchy to the standard NLSE hierarchy. We propose that the dual hierarchy is also integrable, and we confirm this to the first non-trivial order. In the discrete case the requirement of reparametrization invariance is void. But the demand of invariance under local frame rotations prevails, and we utilize it to introduce a discrete va...

  5. Continuous energy cross section library for MCNP/MCNPX based on JENDL high energy file 2007; FXJH7

    OpenAIRE

    佐々 敏信; 菅原 隆徳; 小迫 和明; 深堀 智生

    2008-01-01

    The latest JENDL High Energy File (JENDL/HE) was released in 2007 to respond the requirements of reaction data in high energy range up to several GeV to design accelerator facilities such as accelerator-driven systems and research complex like J-PARC. To apply the JENDL/HE-2007 file to the design study, the cross section library of FXJH7 series was constructed from the JENDL/HE file for the calculation using MCNP and MCNPX codes which are widely used in the field of nuclear reactors, fusion r...

  6. Carbon flow electrodes for continuous operation of capacitive deionization and capacitive mixing energy generation

    NARCIS (Netherlands)

    Porada, S.; Hamelers, H.V.M.; Bryjak, M.; Presser, V.; Biesheuvel, P.M.; Weingarth, D.

    2014-01-01

    Capacitive technologies, such as capacitive deionization and energy harvesting based on mixing energy (“capmix” and “CO2 energy”), are characterized by intermittent operation: phases of ion electrosorption from the water are followed by system regeneration. From a system application point of view, c

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  8. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  10. The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

  11. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

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

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-12-15

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

  14. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

  15. Angra 1 high burnup fuel behaviour under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: dsgomes@ipen.b, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 16x16 NGF (Next Generation Fuel) fuel assembly, comprising of highly corrosive-resistant ZIRLO clad fuel rods, been replacing the current 16x16 Standard (16STD) fuel assembly in the Angra 1, a pressurized water reactor, with a net output of 626 MWe. The 16x16 NGF fuel assemblies are designed for a peak rod average burnup of up to 75 GWd/MTU, thus improving fuel utilization and reducing spent fuel storage issues. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for a further increase in burnup as the simulated RIA tests revealed a lower enthalpy threshold for fuel failure. Two fuel performance codes, FRAPCON and FRAPTRAN, were used to predict high burnup behavior of Angra 1, during an RIA. The maximum average linear fuel rating used was 17.62 KW/m. The FRAPCON 3.4 code was applied to simulate the steady-state performance of the 16 NGF fuel rods up to a burnup of 55 GWd/MTU. With FRAPTRAN-1.4 the fuel behavior was simulated for an RIA power pulse of 4.5 ms (FHWH), and enthalpy peak of 130 Cal/g. With FRAPCON-3.4, the corrosion and hydrogen pickup characteristics of the advanced ZIRLO clad fuel rods were added to the code by modifying the actual corrosion model for Zircaloy-4 through the multiplication of empirical factors, which were appropriate to each alloy, and by means of reducing the current hydrogen pickup fraction. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

    1997-08-01

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

  17. Analysis of the effect of UO{sub 2} high burnup microstructure on fission gas release

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-10-01

    This report deals with high-burnup phenomena with relevance to fission gas release from UO{sub 2} nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  19. Evaluation of the Frequency for Gas Sampling for the High Burnup Confirmatory Data Project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations.

  20. EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

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

    Science.gov (United States)

    Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

    2014-06-01

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

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

    Directory of Open Access Journals (Sweden)

    Lecarpentier D.

    2013-03-01

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

  3. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  4. Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

    Directory of Open Access Journals (Sweden)

    Gajda Paweł

    2016-01-01

    Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

  5. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-01

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

  7. Oxygen potential in the rim region of high burnup UO 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1994-01-01

    Small specimens from the rim region (fuel surface) of a UO 2 fuel rod with an average burnup of 7.6% FIMA were analysed in a miniaturized galvanic cell to determine their oxygen potential Δ Ḡ(O 2) . These fuel pieces represented the porous rim structure with very small grains known to be formed near the periphery of high burnup UO 2 fuel pellets. The oxygen potential of the rim material was very low, corresponding to that of unirradiated stoichiometric UO 2, or to that of slightly substoichiometric UO 2 containing rare earth fission products. No indication of oxidation due to fission was found, though most fission was that of Pu. Measurements on pieces from the inner, unrestructured fuel showed somewhat higher oxygen potentials corresponding to those of very slightly substoichiometric fuel if allowance is made for the incorporation of rare earths. These results are in contrast to some generally accepted ideas of burnup effects, and the possible reasons and implications are discussed.

  8. Estimating Energy- and Eco-Balances for Continuous Bio-Ethanol Production Using a Blenke Cascade System

    Directory of Open Access Journals (Sweden)

    Reinhard Kohlus

    2013-04-01

    Full Text Available Energy and environmental effects of wheat-based fuel, produced continuously by a Blenke cascade system, were assessed. Two scenarios: (1 no-co-products utilization scenario; and (2 co-products utilization scenario, were compared. A Life Cycle Assessment (LCA model was used for analysis. The scope covered a cradle-to-gate inventory. The results from energy analysis showed, that wheat-based ethanol has a positive average net energy value (NEV, NEV = 3.35 MJ/kg ethanol with an average net energy ratio (NER, NER = 1.14 MJ/MJ fossils for scenario 1, while for scenario 2, NEV = 20 MJ/kg ethanol with NER = 3.94 MJ/MJ fossils. The environmental performance analysis indicated that in scenario 1, the strongest contribution to environmental impacts was from the ethanol conversion stage; whereas in scenario 2, it was from wheat production stage. The use of a continuous fermentation system based on Blenke cascade is a promising technology that increases wheat based bio-ethanol’s energy benefits. In addition, the calculated parameters show the potential to significantly reduce emissions level.

  9. Estimating Energy- and Eco-Balances for Continuous Bio-Ethanol Production Using a Blenke Cascade System

    Energy Technology Data Exchange (ETDEWEB)

    Ntihuga, Jean Nepomuscene [Department of Fermentation Technology, Institute of Food Science and Biotechnology, Hohenheim Univ., Stuttgart (Germany) and Department of Food Process Engineering, Institute of Food Science and Biotechnology, Hohenheim Univ., Stuttgart (Germany); Senn, Thomas [Department of Fermentation Technology, Institute of Food Science and Biotechnology, Hohenheim Univ., Stuttgart (Germany); Gschwind, Peter [Department of Food Process Engineering, Institute of Food Science and Biotechnology, Hohenheim Univ., Stuttgart (Germany); Kohlus, Reinhold [Department of Food Process Engineering, Institute of Food Science and Biotechnology, Hohenheim Univ., Stuttgart (Germany)

    2013-04-15

    Energy and environmental effects of wheat-based fuel, produced continuously by a Blenke cascade system, were assessed. Two scenarios: (1) no-co-products utilization scenario; and (2) co-products utilization scenario, were compared. A Life Cycle Assessment (LCA) model was used for analysis. The scope covered a cradle-to-gate inventory. The results from energy analysis showed, that wheat-based ethanol has a positive average net energy value (NEV), NEV = 3.35 MJ/kg ethanol with an average net energy ratio (NER), NER = 1.14 MJ/MJ fossils for scenario 1, while for scenario 2, NEV = 20 MJ/kg ethanol with NER = 3.94 MJ/MJ fossils. The environmental performance analysis indicated that in scenario 1, the strongest contribution to environmental impacts was from the ethanol conversion stage; whereas in scenario 2, it was from wheat production stage. The use of a continuous fermentation system based on Blenke cascade is a promising technology that increases wheat based bio-ethanol’s energy benefits. In addition, the calculated parameters show the potential to significantly reduce emission levels.

  10. A technique for continuous bedside monitoring of global cerebral energy state

    DEFF Research Database (Denmark)

    Jakobsen, Rasmus; Halfeld Nielsen, Troels; Granfeldt, Asger;

    2016-01-01

    BACKGROUND: Cerebral cytoplasmatic redox state is a sensitive indicator of cerebral oxidative metabolism and is conventionally evaluated from the extracellular lactate/pyruvate (LP) ratio. In the present experimental study of global cerebral ischemia induced by hemorrhagic shock, we investigate...... whether the LP ratio obtained from microdialysis of cerebral venous blood may be used as a surrogate marker of global cerebral energy state. METHODS: Six female pigs were anesthetized and vital parameters were recorded. Microdialysis catheters were placed in the left parietal lobe, the superior sagittal...... by severe hemorrhagic shock, intravascular microdialysis of the draining venous blood will exhibit changes of the LP ratio revealing the deterioration of global cerebral oxidative energy metabolism. In neurocritical care, this technique might be used to give information regarding global cerebral energy...

  11. A Leaf-Inspired Luminescent Solar Concentrator for Energy-Efficient Continuous-Flow Photochemistry.

    Science.gov (United States)

    Cambié, Dario; Zhao, Fang; Hessel, Volker; Debije, Michael G; Noël, Timothy

    2017-01-19

    The use of solar light to promote chemical reactions holds significant potential with regard to sustainable energy solutions. While the number of visible light-induced transformations has increased significantly, the use of abundant solar light has been extremely limited. We report a leaf-inspired photomicroreactor that constitutes a merger between luminescent solar concentrators (LSCs) and flow photochemistry to enable green and efficient reactions powered by solar irradiation. This device based on fluorescent dye-doped polydimethylsiloxane collects sunlight, focuses the energy to a narrow wavelength region, and then transports that energy to embedded microchannels where the flowing reactants are converted. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  12. The energy-related inventions program: Continuing benefits to the inventor community

    Energy Technology Data Exchange (ETDEWEB)

    Braid, R.B. Jr.; Brown, M.A.; Wilson, C.R.; Franchuk, C.A.; Rizy, C.G.

    1996-10-01

    This report provides information on the economic, energy, and environmental impacts of inventions supported by the Energy-Related Inventions Program (ERIP) - a technology commercialization program jointly operated by the U.S. Department of Energy (DOE) and the National Institute of Standards and Technology (NIST). It describes the results of the latest in a series of ERIP evaluation projects that have been completed since 1980. The period of interest is 1980 through 1994. The evaluation is based on data collected in 1995 through mail and telephone surveys of 211 program participants, and historical data collected during previous evaluations for an additional 253 participants. As of September 1993, a total of 609 inventions had been recommended to DOE by NIST, which screens all submitted inventions for technical merit, potential for commercial success, and potential energy impact. By the end of 1994, at least 144 (or 24%) of these inventions had entered the market, generating total cumulative sales of $961 million (in 19944). It is estimated that in 1994 ERIP inventors earned royalties of $2.3 million, and over the lifetime of the program, royalties total $28.2 million. With $47.5 million in grants awarded from 1975 through 1994 and $124 million in program appropriations over the same period, ERIP has generated a 20:1 return in terms of sales values to grants, and an 8:1 return in sales versus program appropriations. Further, it is estimated that at least 757 job-years of employment were supported by ERIP technologies in 1994, and that this resulted in a return of approximately $3.4 million in individual income taxes to the U.S. Treasury. Finally, approximately $334 million of energy expenditures were saved in 1994 as a result of the commercial success of five ERIP projects. These energy savings resulted in reduced emissions of 2.1 million metric tons of carbon in 1994 alone.

  13. Analytic continuation and high energy estimates for the resolvent of the Laplacian on forms on asymptotically hyperbolic spaces

    OpenAIRE

    2012-01-01

    We show the analytic continuation of the resolvent of the Laplacian on asymptotically hyperbolic spaces on differential forms, including high energy estimates in strips. This is achieved by placing the spectral family of the Laplacian within the framework developed, and applied to scalar problems, by the author recently, roughly by extending the problem across the boundary of the compactification of the asymptotically hyperbolic space in a suitable manner. The main novelty is that the non-sca...

  14. Continuous Energy Improvement in Motor Driven Systems - A Guidebook for Industry

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert A. McCoy and John G. Douglass

    2014-02-01

    This guidebook provides a step-by-step approach to developing a motor system energy-improvement action plan. An action plan includes which motors should be repaired or replaced with higher efficiency models, recommendations on maintaining a spares inventory, and discussion of improvements in maintenance practices. The guidebook is the successor to DOE’s 1997 Energy Management for Motor Driven Systems. It builds on its predecessor publication by including topics such as power transmission systems and matching driven equipment to process requirements in addition to motors.

  15. Fast continuous energy scan with dynamic coupling of the monochromator and undulator at the DEIMOS beamline.

    Science.gov (United States)

    Joly, L; Otero, E; Choueikani, F; Marteau, F; Chapuis, L; Ohresser, P

    2014-05-01

    In order to improve the efficiency of X-ray absorption data recording, a fast scan method, the Turboscan, has been developed on the DEIMOS beamline at Synchrotron SOLEIL, consisting of a software-synchronized continuous motion of the monochromator and undulator motors. This process suppresses the time loss when waiting for the motors to reach their target positions, as well as software dead-time, while preserving excellent beam characteristics.

  16. A technique for continuous bedside monitoring of global cerebral energy state

    DEFF Research Database (Denmark)

    Jakobsen, Rasmus; Nielsen, Troels Halfeld; Granfeldt, Asger

    2016-01-01

    whether the LP ratio obtained from microdialysis of cerebral venous blood may be used as a surrogate marker of global cerebral energy state. METHODS: Six female pigs were anesthetized and vital parameters were recorded. Microdialysis catheters were placed in the left parietal lobe, the superior sagittal...

  17. Vibration attenuation of a continuous rotor-blisk-journal bearing system employing smooth nonlinear energy sinks

    Science.gov (United States)

    Bab, Saeed; Khadem, S. E.; Shahgholi, Majid; Abbasi, Amirhassan

    2017-02-01

    The current paper investigates the effects of a number of smooth nonlinear energy sinks (NESs) located on the disk and bearings on the vibration attenuation of a rotor-blisk-journal bearing system under excitation of a mass eccentricity force. The blade and rotor are modeled using the Euler-Bernoulli beam theory. The nonlinear energy sinks on the bearing have a linear damping and an essentially nonlinear stiffness. The nonlinear energy sinks on the disk have a linear damping, linear stiffness, and an essentially nonlinear stiffness. It can be seen that the linear stiffness of the NESs on the disk is eliminated by the negative stiffness induced by the centrifugal force, and the collection of the NESs can be tuned to a required rotational speed of the rotor by varying the linear stiffness of the NESs. Furthermore, the remained stiffness of the NESs on the disk after elimination of their linear stiffness, would be essentially a nonlinear (nonlinearizable) one. Two nonlinear energy sinks in the vertical axes are positioned on the bearing housing and nnd NESs are located on the perimeter of the disk. The equations of motion are extracted using the extended Hamilton principle. The modal coordinates and complex transformations are employed to decrease the number of equations of motion. A genetic algorithm is used to optimize the parameters of the nonlinear energy sinks and its objective function is considered as minimizing the vibration of the rotating system within an operating speed range. In order to examine the periodic and non-periodic solutions of the system, time history, bifurcation diagram, Poincaré map, phase portrait, Lyapunov exponent, and power spectra analyses are performed. System shows periodic and quasi-periodic motions for different values of the system parameters. It is shown that the NESs on the disk and bearings have almost local effects on vibration reduction of rotating system. In addition, the optimum NESs remove the instability region from the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-11

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

  19. The 9Be(d,n) 10B-reaction as intense neutron source with continuous energy spectrum

    Science.gov (United States)

    Baumann, F. M.; Domogala, G.; Freiesleben, H.; Paul, H. J.; Puhlvers, S.; Sohlbach, H.

    1986-06-01

    Neutron energy spectra produced by deuterons of 3 to 8 MeV in a thick 9Be-target were measured at various scattering angles. Significant angle dependences were observed. Angular distributions of the most energetic neutrons produced in thin 9Be targets can be described quantitatively in DWBA, which is an indication for a direct reaction mechanism. As a consequence all but 0°-neutrons are polarized to a certain extent. Also presented is the neutron energy spectrum of 7Li(d,n) 8Be at 0° produced in a thick 7Li-target. The potential of these intense 0°-neutron beams with continuous energy distributions is demonstrated by a measurement of the neutron absorption cross section of natural carbon.

  20. Continuous thermal hydrolysis and anaerobic digestion of sludge. Energy integration study.

    Science.gov (United States)

    Pérez-Elvira, S I; Fdz-Polanco, F

    2012-01-01

    Experimental data obtained from the operation in a pilot plant are used to perform mass and energy balances to a global process combining units of thermal hydrolysis (TH) of secondary sludge, anaerobic digestion (AD) of hydrolysed secondary sludge together with fresh primary sludge, and cogeneration from biogas by using a gas engine in which the biogas produces electricity and heat from the exhaust gases. Three scenarios were compared, corresponding to the three digesters operated: C (conventional AD, 17 days residence time), B (combined TH + AD, same time), and A (TH + AD at half residence time). The biogas production of digesters B and A was 33 and 24% better, respectively when compared with C. In the case of the combined TH + AD process (scenarios A and B), the key factors in the energy balance were the recovery of heat from hot streams, and the concentration of sludge. The results of the balances showed that for 8% DS concentration of the secondary sludge tested in the pilot plant, the process can be energetically self-sufficient, but a fraction of the biogas must by-pass the gas engine to be directly burned. From an economic point of view, scenario B is more profitable in terms of green energy and higher waste removal, while scenario A reduces the digester volume required by a half. Considering a population of 100,000 inhabitants, the economic benefit is 87,600 €/yr for scenario A and 132,373 €/yr for B. This value can be increased to 223,867 €/yr by increasing the sludge concentration of the feeding to the TH unit to a minimum value that allows use of all the biogas to produce green energy. This concentration is 13% DS, which is still possible from a practical point of view. Additional benefits gained with the combined TH + AD process are the enhancement of the digesters rheology and the possibility of getting Class A biosolids. The integration study presented here set the basis for the scale-up to a demonstration plant.

  1. From fixed-energy MSA to dynamical localization: A continuing quest for elementary proofs

    CERN Document Server

    Chulaevsky, Victor

    2012-01-01

    We review several techniques and ideas initiated by a remarkable work by Spencer [26], used and further developed in numerous subsequent researches. We also describe a relatively short and elementary derivation of the spectral and strong dynamical Anderson localization from the fixed-energy analysis of the Green functions, obtained either by the Multi-Scale Analysis (MSA) or by the Fractional-Moment Method (FMM). This derivation goes in the same direction as the Simon--Wolf criterion [28], but provides quantitative estimates, applies also to multi-particle models and, combined with a simplified variant of the Germinet--Klein argument [20], results in an elementary proof of dynamical localization.

  2. Reducing temperature dependence of the output energy of a quasi-continuous wave diode-pumped Nd:YAG laser.

    Science.gov (United States)

    Lee, Kangin; Kim, Youngjung; Lee, Sijin; Kwon, Jin Hyuk; Gwak, Jin Seog; Yi, Jonghoon

    2013-08-20

    It is demonstrated by numerical modeling that spectrally dispersed compound pumping diodes and low-loss pumping chamber reduced the temperature dependence of the output energy of quasi-continuous wave diode-pumped Nd:YAG lasers considerably. Several compound diodes with different spectral profiles were tested for pumping. The laser energy was calculated as a function of diode temperature from -30°C to 60°C. When a compound diode with a flat-top spectrum was used for pumping, the mean laser energy was 83% of the maximum energy of a Nd:YAG laser pumped by a diode with a narrow bandwidth. In addition, a compound diode with three emission lines was tested for pumping. When the wavelength gap between the adjacent emission lines of the pumping diode was in the range of 3-10 nm, the mean energy of the Nd:YAG laser became similar to that of a Nd:YAG laser pumped by a diode with a flat-top spectrum.

  3. Energy absorption, lean body mass, and total body fat changes during 5 weeks of continuous bed rest

    Science.gov (United States)

    Krebs, Jean M.; Evans, Harlan; Kuo, Mike C.; Schneider, Victor S.; Leblanc, Adrian D.

    1990-01-01

    The nature of the body composition changes due to inactivity was examined together with the question of whether these changes are secondary to changes in energy absorption. Volunteers were 15 healthy males who lived on a metabolic research ward under close staff supervision for 11 weeks. Subjects were ambulatory during the first six weeks and remained in continuous bed rest for the last five weeks of the study. Six male volunteers (age 24-61 years) were selected for body composition measurements. Nine different male volunteers (age 21-50 years) were selected for energy absorption measurements. The volunteers were fed weighed conventional foods on a constant 7-d rotation menu. The average daily caloric content was 2,592 kcal. Comparing the five weeks of continuous bed rest with the previous six weeks of ambulation, it was observed that there was no change in energy absorption or total body weight during bed rest, but a significant decrease in lean body mass and a significant increase in total body fat (p less than 0.05).

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

    Energy Technology Data Exchange (ETDEWEB)

    Chon, Je Keun; Kim, Jae Chun; Koh, Duck Joon; Kim Byung Tae [Nuclear Environment Technology Institute, Korea Electric Power Corporation, Taejon (Korea, Republic of)

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  6. Spent fuel dissolution rates as a function of burnup and water chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  7. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  8. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-30

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

  10. Thermodynamic analysis for high burn-up fuel internal chemistry. 2

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan)

    1998-09-01

    Thermodynamic calculations with the computer program SOLGASMIX-PV have been performed for the chemical states expected in irradiated fast breeder reactor (FBR) fuels containing transuranium (TRU) elements. The analysis shows that A (alkali and alkaline-earth)-molybdates exist, but neither A-uranates nor A-zirconates are formed in FBR fuel pellets irradiated to high burn-up. And increase of oxygen potential in irradiated FBR fuel is ascribed to growing amount of rare earth, noble metal and TRU elements. (author)

  11. An Aquifer Thermal Energy Storage (ATES) System for Continuous and Sustainable Cold Supply in Oman

    Science.gov (United States)

    Winterleitner, G.; Schütz, F.; Huenges, E.

    2016-12-01

    The aim of the GeoSolCool research programme between the German Research Centre for Geoscience (GFZ) and The Research Council of Oman (TRC) is the development of an innovative and sustainable cooling system in combination with an aquifer thermal energy storage system in northern Oman. An integral part of this project is the design of a subsurface aquifer reservoir system for storage of thermal energy through hot water injection. An accurate characterisation of potential storage horizons is thus essential to ensure optimal efficiency of the cooling system. The study area, 40 km west of Muscat is characterised by a thick Cenozoic mixed carbonate-siliciclastic sedimentary succession, containing at least 3 aquifer horizons. We used a multidisciplinary approach for the initial ATES development phase, including geological fieldwork dovetailed with remote sensing analyses, thin-section analyses, geological modelling and reservoir fluid flow forecasting. First results indicate two potential storage horizons: (1) a Miocene-aged clastic-dominated alluvial fan system and (2) an Eocene carbonate sequence. The alluvial fan system is a more than 300 m thick, coarse clastic (mainly gravels and sandstones) succession of coalesced individual fans. Thin-section analyses showed that hydraulic parameters are favourable for the gravel and sandstone intervals but reservoir architecture is complex due to multiple generations of interconnecting fans with highly heterogeneous facies distributions. The Eocene carbonates were deposited in a carbonate ramp setting, strongly influenced by currents and storm events. Individual facies belts extend over kilometres and thus horizontal reservoir connectivity is expected to be good with minor facies variability. Thin-section analyses showed that especially the fossil-rich sections show good storage qualities. Fluid flow forecasting indicate that both potential horizons have good to very good storage characteristics. However, intense diagenetic

  12. Field test of an all-semiconductor laser-based coherent continuous-wave Doppler lidar for wind energy applications

    DEFF Research Database (Denmark)

    Sjöholm, Mikael; Dellwik, Ebba; Hu, Qi

    The wind energy industry is gaining interest in prevision of the rotor inflow for turbine control. The potential benefits are increased power production due to better alignment of the rotor to the mean wind direction as well as prolonged lifetime of the turbine due to load reductions. Several lidar......-produced all-semiconductor laser. The instrument is a coherent continuous-wave lidar with two fixed-focus telescopes for launching laser beams in two different directions. The alternation between the telescopes is achieved by a novel switching technique without any moving parts. Here, we report results from...

  13. Research and verification of Monte Carlo burnup calculations based on Chebyshev rational approximation method%基于切比雪夫有理逼近方法的蒙特卡罗燃耗计算研究与验证

    Institute of Scientific and Technical Information of China (English)

    范文玎; 孙光耀; 张彬航; 陈锐; 郝丽娟

    2016-01-01

    燃耗计算在反应堆设计、分析研究中起着重要作用.相比于传统点燃耗算法,切比雪夫有理逼近方法(Chebyshev rational approximation method,CRAM)具有计算速度快、精度高的优点.基于超级蒙特卡罗核计算仿真软件系统SuperMC(Super Monte Carlo Simulation Program for Nuclear and Radiation Process),采用切比雪夫有理逼近方法和桶排序能量查找方法,进行了蒙特卡罗燃耗计算的初步研究与验证.通过燃料棒燃耗例题以及IAEA-ADS(International Atomic Energy Agency-Accelerator Driven Systems)国际基准题,初步验证了该燃耗计算方法的正确性,且IAEA-ADS基准题测试表明,与统一能量网格方法相比,桶排序能量查找方法在保证了计算效率的同时减少了内存开销.%Background:Burnup calculation is the key point of reactor design and analysis. It's significant to calculate the burnup situation and isotopic atom density accurately while a reactor is being designed.Purpose:Based on the Monte Carlo particle simulation code SuperMC (Super Monte Carlo Simulation Program for Nuclear and Radiation Process), this paper aimed to conduct preliminary study and verification on Monte Carlo burnup calculations. Methods:For the characteristics of accuracy, this paper adopted Chebyshev rational approximation method (CRAM) as the point-burnup algorithm. Moreover, instead of the union energy grids method, this paper adopted an energy searching method based on bucket sort algorithm, which reduced the memory overhead on the condition that the calculation efficiency is ensured.Results:By calculating the fuel rod burnup problem and the IAEA-ADS (International Atomic Energy Agency - Accelerator Driven Systems) international benchmark, the simulation results were basically consistent with Serpent and other counties' results, respectively. In addition, the bucket sort energy searching method reduced about 95% storage space compared with union energy grids method for IAEA

  14. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    Science.gov (United States)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  15. A continuous latitudinal energy balance model to explore non-uniform climate engineering strategies

    Science.gov (United States)

    Bonetti, F.; McInnes, C. R.

    2016-12-01

    Current concentrations of atmospheric CO2 exceed measured historical levels in modern times, largely attributed to anthropogenic forcing since the industrial revolution. The required decline in emissions rates has never been achieved leading to recent interest in climate engineering for future risk-mitigation strategies. Climate engineering aims to offset human-driven climate change. It involves techniques developed both to reduce the concentration of CO2 in the atmosphere (Carbon Dioxide Removal (CDR) methods) and to counteract the radiative forcing that it generates (Solar Radiation Management (SRM) methods). In order to investigate effects of SRM technologies for climate engineering, an analytical model describing the main dynamics of the Earth's climate has been developed. The model is a time-dependent Energy Balance Model (EBM) with latitudinal resolution and allows for the evaluation of non-uniform climate engineering strategies. A significant disadvantage of climate engineering techniques involving the management of solar radiation is regional disparities in cooling. This model offers an analytical approach to design multi-objective strategies that counteract climate change on a regional basis: for example, to cool the Artic and restrict undesired impacts at mid-latitudes, or to control the equator-to-pole temperature gradient. Using the Green's function approach the resulting partial differential equation allows for the computation of the surface temperature as a function of time and latitude when a 1% per year increase in the CO2 concentration is considered. After the validation of the model through comparisons with high fidelity numerical models, it will be used to explore strategies for the injection of the aerosol precursors in the stratosphere. In particular, the model involves detailed description of the optical properties of the particles, the wash-out dynamics and the estimation of the radiative cooling they can generate.

  16. Mechanistic modelling of infrared mediated energy transfer during the primary drying step of a continuous freeze-drying process.

    Science.gov (United States)

    Van Bockstal, Pieter-Jan; Mortier, Séverine Thérèse F C; De Meyer, Laurens; Corver, Jos; Vervaet, Chris; Nopens, Ingmar; De Beer, Thomas

    2017-01-12

    Conventional pharmaceutical freeze-drying is an inefficient and expensive batch-wise process, associated with several disadvantages leading to an uncontrolled end product variability. The proposed continuous alternative, based on spinning the vials during freezing and on optimal energy supply during drying, strongly increases process efficiency and improves product quality (uniformity). The heat transfer during continuous drying of the spin frozen vials is provided via non-contact infrared (IR) radiation. The energy transfer to the spin frozen vials should be optimised to maximise the drying efficiency while avoiding cake collapse. Therefore, a mechanistic model was developed which allows computing the optimal, dynamic IR heater temperature in function of the primary drying progress and which, hence, also allows predicting the primary drying endpoint based on the applied dynamic IR heater temperature. The model was validated by drying spin frozen vials containing the model formulation (3.9mL in 10R vials) according to the computed IR heater temperature profile. In total, 6 validation experiments were conducted. The primary drying endpoint was experimentally determined via in-line near-infrared (NIR) spectroscopy and compared with the endpoint predicted by the model (50min). The mean ratio of the experimental drying time to the predicted value was 0.91, indicating a good agreement between the model predictions and the experimental data. The end product had an elegant product appearance (visual inspection) and an acceptable residual moisture content (Karl Fischer).

  17. Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

    Science.gov (United States)

    Ramamoorthy, Karthikeyan

    The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant

  18. MC21 v.6.0 - A Continuous-Energy Monte Carlo Particle Transport Code with Integrated Reactor Feedback Capabilities

    Science.gov (United States)

    Griesheimer, D. P.; Gill, D. F.; Nease, B. R.; Sutton, T. M.; Stedry, M. H.; Dobreff, P. S.; Carpenter, D. C.; Trumbull, T. H.; Caro, E.; Joo, H.; Millman, D. L.

    2014-06-01

    MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10-5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each

  19. EBSD and TEM characterization of high burn-up mixed oxide fuel

    Science.gov (United States)

    Teague, Melissa; Gorman, Brian; Miller, Brandon; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to ∼1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had ∼2.5× higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice ∼25 μm cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

  20. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  1. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    Science.gov (United States)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  2. Comparison of neutron cross sections for selected fission products and isotopic composition analyses with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Gil, C. S.; Kim, J. D.; Jang, J. H.; Lee, Y. D. [KAERI, Taejon (Korea)

    2003-10-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI-BNL international collaboration have been compared with the ENDF/B-VI release 7. Also, the influence of the new evaluations on isotopic compositions of the fission products as a function of burnup has been analyzed through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69 group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including new evaluations in resonance region covering thermal region, and ENDF/B-VII expected including those in upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows maximum difference of 4.78% compared to ENDF/B-VI.7. However, the isotopic compositions of all fission products calculated with ENDF/B-VII shows no differences compared to ENDF/B-VI.7.

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

    Science.gov (United States)

    Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

    2012-08-01

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

  4. Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2014-01-01

    Full Text Available In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium. Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt neutron generation time L in the understudy core is lower than reference operating core of reactor at all burn-up steps due to hard spectrum. It is observed that beff is larger in the understudy core than reference operating core of due to smaller size. Calculations were performed with the help of computer codes WIMSD/4 and CITATION.

  5. Conceptual design study on an upgraded future Monju core (2). Core concept with extended refueling interval and increased fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kinjo, Hidehito; Ishibashi, Jun-ichi; Nishi, Hiroshi [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Kageyama, Takeshi [Nuclear Energy System Inc., Tokyo (Japan)

    2003-03-01

    A conceptual design study has been performed at the International Cooperation and Technology Development Center to investigate the feasibility of upgraded future Monju cores with extended refueling intervals of 365efpd/cycle and increased fuel burnup of 150 GWd/t. The goal of this study is to demonstrate the possible contribution of Monju to the improved economy and to efficient utilization, as one of the major facilities for fast neutron irradiation. Two design measures have been mainly taken to improve the core fuel burnup and reactivity control characteristics for the extended operating cycle length of 1 year: (1) The driver fuel pin specification with both increased pin diameter of 7.7mm and increased active core height of about 100cm has been chosen to reduce the burnup reactivity swing, (2) The absorber control rod specification has also been changed to enhance the control rod reactivity worth by increasing {sup 10}B-enrichment and absorber length, and to adequately secure the shutdown reactivity margin. The major core characteristics have been evaluated on the core power distribution, safety parameters such as sodium void reactivity and Doppler effect, thermal hydraulics and reactivity control characteristics. The results show that this core could achieve the targeted core performances of 1-year operating cycle as well as 150GWd/t discharged burnup, without causing any significant drawback on the core characteristics and safety aspects. The upgraded core concepts have, therefore, been confirmed as feasible. (author)

  6. Analytical continuation in coupling constant method; application to the calculation of resonance energies and widths for organic molecules: Glycine, alanine and valine and dimer of formic acid

    Science.gov (United States)

    Papp, P.; Matejčík, Š.; Mach, P.; Urban, J.; Paidarová, I.; Horáček, J.

    2013-06-01

    The method of analytic continuation in the coupling constant (ACCC) in combination with use of the statistical Padé approximation is applied to the determination of resonance energy and width of some amino acids and formic acid dimer. Standard quantum chemistry codes provide accurate data which can be used for analytic continuation in the coupling constant to obtain the resonance energy and width of organic molecules with a good accuracy. The obtained results are compared with the existing experimental ones.

  7. Nanoengineering of Ruthenium and Platinum-based Nanocatalysts by Continuous-Flow Chemistry for Renewable Energy Applications

    KAUST Repository

    AlYami, Noktan Mohammed

    2017-04-15

    This thesis presents an integrated study of nanocatalysts for heterogenous catalytic and electrochemical processes using pure ruthenium (Ru) with mixed-phase and platinum-based nanomaterials synthesized by continuous-flow chemistry. There are three major challenges to the application of nanomaterials in heterogenous catalytic reactions and electrocatalytic processes in acidic solution. These challenges are the following: (i) controlling the size, shape and crystallography of nanoparticles to give the best catalytic properties, (ii) scaling these nanoparticles up to a commercial quantity (kg per day) and (iii) making stable nanoparticles that can be used catalytically without degrading in acidic electrolytes. Some crucial limitations of these nanostructured materials in energy conversion and storage applications were overcome by continuous-flow chemistry. By using a continuous-flow reactor, the creation of scalable nanoparticle systems was achieved and their functionality was modified to control the nanoparticles’ physical and chemical characteristics. The nanoparticles were also tested for long-term stability, to make sure these nanoparticles were feasible under realistic working conditions. These nanoparticles are (1) shape- and crystallography-controlled ruthenium (Ru) nanoparticles, (2) size-controlled platinum-metal (Pt-M= nickel (Ni) & copper (Cu)) nanooctahedra (while maintaining morphology) and (3) core-shell platinum@ruthenium (Pt@Ru) nanoparticles where an ultrathin ruthenium shell was templated onto the platinum core. Thus, a complete experimental validation of the formation of a scalable amount of these nanoparticles and their catalytic activity and stability towards the oxygen evolution reaction (OER) in acid medium, hydrolysis of ammonia borane (AB) along with plausible explanations were provided.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  9. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

  10. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    CERN Document Server

    Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

    2014-01-01

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

  11. Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boafo, Emmanuel [Ghana atomic Energy Commission, Accra (Ghana)

    2013-07-01

    The burnup capability in the MCNPX code was utilized to perform fuel depletion analysis of the MNSR LEU core by estimating the amount of fissile material (U-235) consumed as well as the amount of plutonium formed after the reactor core expected life. The decay heat removal rate for the MNSR after reactor shutdown was also investigated due to its significance to reactor safety. The results show that 0.568 % of U-235 was burnt up after 200 days of reactor operation while the amount of plutonium formed was not significant. The study also found that the decay heat decreased exponentially after reactor shutdown confirming that the decay heat will be removed from the system by natural circulation after shut down and hence safety of the reactor is assured.

  12. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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).

  15. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    Science.gov (United States)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  16. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  17. Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis

    Science.gov (United States)

    Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.

    2012-12-01

    The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant α. We have found out, that for the rate of α change, the inequality ěrt˙ {α }/α ěrt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.

  18. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    Science.gov (United States)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  19. Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis

    CERN Document Server

    Onegin, M S

    2010-01-01

    The burn-up for SC56-1472 sample of the natural Oklo reactor zone 3 was calculated using the modern Monte Carlo codes. We reconstructed the neutron spectrum in the core by means of the isotope ratios: $^{147}$Sm/$^{148}$Sm and $^{176}$Lu/$^{175}$Lu. These ratios unambiguously determine the spectrum index and core temperature. The effective neutron absorption cross section of $^{149}$Sm calculated using this spectrum was compared with experimental one. The disagreement between these two values allows to limit a possible shift of the low laying resonance of $^{149}$Sm even more . Then, these limits were converted to the limits for the change of the fine structure constant $\\alpha$. We found that for the rate of $\\alpha$ change the inequality $|\\delta \\dot{\\alpha}/\\alpha| \\le 5\\cdot 10^{-18}$ is fulfilled, which is of the next higher order than our previous limit.

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

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

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

  1. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Betzler, Benjamin R [ORNL; Ade, Brian J [ORNL

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-17

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

  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. Annual Continuation And Progress Report For Low-Energy Nuclear Physics Research At Lawrence Livermore National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Scielzo, N. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wu, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-27

    Ba where the octupole deformation is evident from the measured B(E3; 3-→0+) strengths that significantly greater than the theoretical predictions. We anticipate that CHICO2 will continue to be a viable charged-­particle detector for the research need of the low-­energy nuclear physics community.

  5. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, T.A.

    1980-03-01

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation.

  6. Design of an 81.25 MHz continuous-wave radio-frequency quadrupole accelerator for Low Energy Accelerator Facility

    Science.gov (United States)

    Ma, Wei; Lu, Liang; Xu, Xianbo; Sun, Liepeng; Zhang, Zhouli; Dou, Weiping; Li, Chenxing; Shi, Longbo; He, Yuan; Zhao, Hongwei

    2017-03-01

    An 81.25 MHz continuous wave (CW) radio frequency quadrupole (RFQ) accelerator has been designed for the Low Energy Accelerator Facility (LEAF) at the Institute of Modern Physics (IMP) of the Chinese Academy of Science (CAS). In the CW operating mode, the proposed RFQ design adopted the conventional four-vane structure. The main design goals are providing high shunt impendence with low power losses. In the electromagnetic (EM) design, the π-mode stabilizing loops (PISLs) were optimized to produce a good mode separation. The tuners were also designed and optimized to tune the frequency and field flatness of the operating mode. The vane undercuts were optimized to provide a flat field along the RFQ cavity. Additionally, a full length model with modulations was set up for the final EM simulations. Following the EM design, thermal analysis of the structure was carried out. In this paper, detailed EM design and thermal simulations of the LEAF-RFQ will be presented and discussed. Structure error analysis was also studied.

  7. Burnup extension and evolution in the fuel management of EDF's nuclear power plants; Accroissement des taux de combustion et impact des evolutions de gestion sur l'exploitation des reacteurs du parc EDF

    Energy Technology Data Exchange (ETDEWEB)

    Provost, J.L.; Thibault, X. [Electricite de France (EDF/DPN), 93 - Saint-Denis (France); Debes, M. [Electricite de France (EDF/DCN), 92 - Clamart (France); Kaplan, P. [Cogema, 78 - Velizy Villacoublay (France)

    2004-07-01

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

  8. Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

    OpenAIRE

    岡野 正紀; 久野 剛彦; 高橋 一朗; 白水 秀知; Charlton, W. S.; Wells, C. A.; Hemberger, P. H.; 山田 敬二; 酒井 敏雄

    2006-01-01

    The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant during BWR fuel reprocessing campaign. Xenon isotopic ratio was determined with GC/MS. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Labo...

  9. Analytical continuation in coupling constant method; application to the calculation of resonance energies and widths for organic molecules: Glycine, alanine and valine and dimer of formic acid

    Energy Technology Data Exchange (ETDEWEB)

    Papp, P., E-mail: papp@fmph.uniba.sk [Department of Experimental Physics, Faculty of Mathematics, Physics and Informatics, Comenius University, Mlynská dolina, 84248 Bratislava (Slovakia); Matejčík, Š. [Department of Experimental Physics, Faculty of Mathematics, Physics and Informatics, Comenius University, Mlynská dolina, 84248 Bratislava (Slovakia); Mach, P.; Urban, J. [Department of Nuclear Physics and Biophysics, Faculty of Mathematics, Physics and Informatics, Comenius University, Mlynská dolina, 84248 Bratislava (Slovakia); Paidarová, I. [J. Heyrovský Institute of Physical Chemistry, Academy of Sciences of the Czech Republic, v.v.i., Dolejškova 3, CZ-182 23 Praha 8 (Czech Republic); Horáček, J., E-mail: horacek@mbox.troja.mff.cuni.cz [Charles University, Faculty of Mathematics and Physics, V Holešovičkách 2, CZ-180 00 Praha 8 (Czech Republic)

    2013-06-03

    Highlights: • The anions are stabilized by additional charges on the nuclei. • The energy dependence of anions and neutrals on nuclear charges are calculated by ab initio methods. • Resonance energies and widths are obtained from the energy data by analytical continuation with Padé approximation. • The resonance energies and widths of amino acids are compared with Nestmann–Peyerimhoff’s method and with experiment. • The resonance energies and (widths) of formic acid monomer and dimer are 2.09 (0.33) eV and 1.7 (0.13) eV, respectively. - Abstract: The method of analytic continuation in the coupling constant (ACCC) in combination with use of the statistical Padé approximation is applied to the determination of resonance energy and width of some amino acids and formic acid dimer. Standard quantum chemistry codes provide accurate data which can be used for analytic continuation in the coupling constant to obtain the resonance energy and width of organic molecules with a good accuracy. The obtained results are compared with the existing experimental ones.

  10. Effects of continuous venovenous haemofiltration-induced cooling on global haemodynamics, splanchnic oxygen and energy balance in critically ill patients.

    Science.gov (United States)

    Rokyta, Richard; Matejovic, Martin; Krouzecky, Ales; Opatrny, Karel; Ruzicka, Jiri; Novak, Ivan

    2004-03-01

    A number of haemodialysis studies have demonstrated beneficial effects of cooler dialysates on global haemodynamics in chronic dialysis patients. However, the effects of continuous venovenous haemofiltration (CVVH)-induced cooling on regional perfusion and energy metabolism in critically ill septic patients have not been well defined. Nine septic mechanically ventilated patients (age 40-69 years) were investigated during CVVH (ultrafiltration 30-35 ml/kg/h). Baseline data (=WARM 1) were collected when core temperature (Tc) was >37.5 degrees C; the second data set (=COLD) was obtained after 120 min of 'cooling'; and a third set (=WARM 2) was obtained after 120 min of 'rewarming'. During 'warming' (WARM 1 and 2, respectively), both substitution fluids (SFs) and 'returned' blood (RB) were warmed (37 degrees C), whereas during 'cooling', the SFs were at 20 degrees C and RB was not warmed. We measured hepatic venous (HV) haemoglobin oxygen saturation (ShvO(2)), blood gases, lactate and pyruvate. Gastric mucosal PCO(2) (PgmCO(2)) was measured by air tonometry and the gastric mucosal - arterial PCO(2) difference (PCO(2) gap) was calculated. Haemodynamic monitoring was performed with arterial and pulmonary arterial thermodilution catheters. Tcs were significantly altered [WARM 1, 37.9 degrees C (37.6, 38.3); COLD, 36.8 degrees C (36.3, 37.1); WARM 2, 37.5 degrees C (37.0, 38.0); Pcooling. As a result, mean arterial pressure increased. Cooling was associated with significant decreases in heart rate, cardiac output, systemic oxygen delivery and consumption. ShvO(2) did not change [WARM 1, 51.0% (44.0, 59.5); COLD, 49.0% (42.0, 58.0); WARM 2, 51.0% (46.0, 57.0); P = NS]. The splanchnic oxygen extraction ratio, the HV lactate to pyruvate ratio, HV acid base status and PCO(2) gap remained unchanged. Mild core cooling induced by CVVH may not affect hepatosplanchnic oxygen and energy balance in septic critically ill patients, even though it affects global haemodynamics.

  11. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  12. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    Energy Technology Data Exchange (ETDEWEB)

    Sataka, M.; Ishikawa, N.; Chimn, Y.; Nakamura, J.; Amaya, M. [Japan Atomic Energy Agency, Naka Gun (Japan); Iwasawa, M.; Ohnuma, T.; Sonoda, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Kinoshita, M.; Geng, H. Y.; Chen, Y.; Kaneta, Y. [The Univ. of Tokyo, Tokyo (Japan); Yasunaga, K.; Matsumura, S.; Yasuda, K. [Kyushu Univ., Motooka (Japan); Iwase [Osaka Prefecture Univ., Osaka (Japan); Ichinomiya, T.; Nishiuran, Y. [Hokkaido Univ., Kitaku (Japan); Matzke, HJ. [Academy of Ceramics, Karlsruhe (Germany)

    2008-10-15

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring.

  13. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    Science.gov (United States)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  14. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

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

    Science.gov (United States)

    Carmack, William J.

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

  16. A feasibility study to determine cooling time and burnup of ATR fuel using a nondestructive technique and three types of gamma-ray detectors

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, J.; Aryaeinejad, R.; Nigg, D.W. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr{sub 3}), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry. (authors)

  17. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

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

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E., E-mail: sskutnik@utk.edu; Davis, David R.

    2016-05-01

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

  1. 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)

  2. 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.

  3. MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP Obrigheim

    Energy Technology Data Exchange (ETDEWEB)

    Cao Yan, E-mail: ycao@anl.go [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Broeders, Cornelis H.M. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2010-10-15

    This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for {approx}10% differences in the prediction of the minor actinide isotopes buildup.

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

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

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

  5. Development and verification of fuel burn-up calculation model in a reduced reactor geometry

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, Tagor Malem [Center for Reactor Technology and Nuclear Safety (PTKRN), National Nuclear Energy Agency (BATAN), Kawasan PUSPIPTEK Gd. No. 80, Serpong, Tangerang 15310 (Indonesia)], E-mail: tagorms@batan.go.id; Liem, Peng Hong [Research Laboratory for Nuclear Reactor (RLNR), Tokyo Institute of Technology (Tokyo Tech), O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2008-02-15

    A fuel burn-up model in a reduced reactor geometry (2-D) is successfully developed and implemented in the Batan in-core fuel management code, Batan-FUEL. Considering the bank mode operation of the control rods, several interpolation functions are investigated which best approximate the 3-D fuel assembly radial power distributions across the core as function of insertion depth of the control rods. Concerning the applicability of the interpolation functions, it can be concluded that the optimal coefficients of the interpolation functions are not very sensitive to the core configuration and core or fuel composition in RSG GAS (MPR-30) reactor. Consequently, once the optimal interpolation function and its coefficients are derived then they can be used for 2-D routine operational in-core fuel management without repeating the expensive 3-D neutron diffusion calculations. At the selected fuel elements (at H-9 and G-6 core grid positions), the discrepancy of the FECFs (fuel element channel power peaking factors) between the 2-D and 3-D models are within the range of 3.637 x 10{sup -4}, 3.241 x 10{sup -4} and 7.556 x 10{sup -4} for the oxide, silicide cores with 250 g {sup 235}U/FE and the silicide core with 300 g {sup 235}U/FE, respectively.

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

    Science.gov (United States)

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

    2014-04-01

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

  7. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    CERN Document Server

    Chiesa, Davide; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2015-01-01

    A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate experimental reactors from power ones, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the...

  8. First steps towards a validation of the new burnup and depletion code TNT

    Energy Technology Data Exchange (ETDEWEB)

    Herber, S.C.; Allelein, H.J. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6); Friege, N. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Kasselmann, S. [Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6)

    2012-11-01

    In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)

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

    Science.gov (United States)

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

    2015-09-01

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

  10. SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

    2014-04-01

    During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.

  11. Irradiation characteristics examination technology development of irradiated nuclear material and high burn-up fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Kwon Pyo; Choo, Y. S.; Oh, Y. W. [and others

    2002-12-01

    The research and development for the first year of the project are performed through specialization of researchers, information from aborad and international cooperation, securement of advanced nuclear technology, development and installation of test equipment, application of external man-power, establishment of advanced test techniques, and certified test method. 1. Absolute efficiency measurement examination technology development of gamma scanning system 2. Sample preparation technology development of SEM and EPMA for micro-structural observation and chemical composition analysis 3. Irradiated high burn-up nuclear fuel transportation and test for PWR 4. Development of hot cell examination techniques and equipment 5. Acquirement of KOLAS system. In addition to the project, the following activities are carried out as follows; - PIE of Hanaro fuel(KH99H-001) - PIE of U-Mo advanced nuclear fuel irradiated at Hanaro - PIE of Hi-MET advanced nuclear fuel irradiated at Hanaro - PIE of DUPIC project - Hot cell examination of Hanaro irradiated capsule - Leaching test of PWR fuels - Surveillance test of PWR vessels - Mechanical test of CANDU pressure tubes.

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

    CERN Document Server

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

    2013-01-01

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

  13. TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

    2014-04-01

    As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

  14. Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

    Science.gov (United States)

    Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

    2013-09-01

    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 UO2 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 UO2. 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 UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 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 UO2 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 UO2. 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.

  15. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  16. Continuous and low-energy 125I seed irradiation changes DNA methyltransferases expression patterns and inhibits pancreatic cancer tumor growth

    Directory of Open Access Journals (Sweden)

    Gong Yan-fang

    2011-04-01

    Full Text Available Abstract Background Iodine 125 (125I seed irradiation is an effective treatment for unresectable pancreatic cancers. However, the radiobiological mechanisms underlying brachytherapy remain unclear. Therefore, we investigated the influence of continuous and low-energy 125I irradiation on apoptosis, expression of DNA methyltransferases (DNMTs and cell growth in pancreatic cancers. Materials and methods For in vitro 125I seed irradiation, SW-1990 cells were divided into three groups: control (0 Gy, 2 Gy, and 4 Gy. To create an animal model of pancreatic cancer, the SW 1990 cells were surgically implanted into the mouse pancreas. At 10 d post-implantation, the 30 mice with pancreatic cancer underwent 125I seed implantation and were separated into three groups: 0 Gy, 2 Gy, and 4 Gy group. At 48 or 72 h after irradiation, apoptosis was detected by flow cytometry; changes in DNMTs mRNA and protein expression were assessed by real-time PCR and western blotting analysis, respectively. At 28 d after 125I seed implantation, in vivo apoptosis was evaluated with TUNEL staining, while DNMTs protein expression was detected with immunohistochemical staining. The tumor volume was measured 0 and 28 d after 125I seed implantation. Results 125I seed irradiation induced significant apoptosis, especially at 4 Gy. DNMT1 and DNMT3b mRNA and protein expression were substantially higher in the 2 Gy group than in the control group. Conversely, the 4 Gy cell group exhibited significantly decreased DNMT3b mRNA and protein expression relative to the control group. There were substantially more TUNEL positive in the 125I seed implantation treatment group than in the control group, especially at 4 Gy. The 4 Gy seed implantation group showed weaker staining for DNMT1 and DNMT3b protein relative to the control group. Consequently, 125I seed implantation inhibited cancer growth and reduced cancer volume. Conclusion 125I seed implantation kills pancreatic cancer cells, especially

  17. 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

  18. Source Term Analysis for Reactor Coolant System with Consideration of Fuel Burnup

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yu Jong; Ahn, Joon Gi; Hwang, Hae Ryong [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    The radiation source terms in reactor coolant system (RCS) of pressurized water reactor (PWR) are basic design information for ALARA design such as radiation protection and shielding. Usually engineering companies own self-developed computer codes to estimate the source terms in RCS. DAMSAM and FIPCO are the codes developed by engineering companies. KEPCO E and C has developed computer code, RadSTAR, for use in the Radiation Source Term Analysis for Reactor coolant system during normal operation. The characteristics of RadSTAR are as follows. (1) RadSTAR uses fuel inventory data calculated by ORIGEN, such as ORIGEN2 or ORIGEN-S to consider effects of the fuel burnup. (2) RadSTAR estimates fission products by using finite differential method and analytic method to minimize numerical error. (3) RadSTAR enhances flexibility by adding the function to build the nuclide data library (production pathway library) for user-defined nuclides from ORIGEN data library. (4) RadSTAR consists of two modules. RadSTAR-BL is to build the nuclide data library. RadSTAR-ST is to perform numerical analysis on source terms. This paper includes descriptions on the numerical model, the buildup of nuclide data library, and the sensitivity analysis and verification of RadSTAR. KEPCO E and C developed RadSTAR to calculate source terms in RCS during normal operation. Sensitivity analysis and accuracy verification showed that RadSTAR keeps stability at Δt of 0.1 day and gives more accurate results in comparison with DAMSAM. After development, RadSTAR will replace DAMSAM. The areas, necessary to further development of RadSTAR, are addition of source term calculations for activation products and for shutdown operation.

  19. COMPARISONS OF THE FINITE-ELEMENT-WITH-DISCONTIGUOUS-SUPPORT METHOD TO CONTINUOUS-ENERGY MONTE CARLO FOR PIN-CELL PROBLEMS

    Energy Technology Data Exchange (ETDEWEB)

    A. T. Till; M. Hanuš; J. Lou; J. E. Morel; M. L. Adams

    2016-05-01

    The standard multigroup (MG) method for energy discretization of the transport equation can be sensitive to approximations in the weighting spectrum chosen for cross-section averaging. As a result, MG often inaccurately treats important phenomena such as self-shielding variations across a material. From a finite-element viewpoint, MG uses a single fixed basis function (the pre-selected spectrum) within each group, with no mechanism to adapt to local solution behavior. In this work, we introduce the Finite-Element-with-Discontiguous-Support (FEDS) method, whose only approximation with respect to energy is that the angular flux is a linear combination of unknowns multiplied by basis functions. A basis function is non-zero only in the discontiguous set of energy intervals associated with its energy element. Discontiguous energy elements are generalizations of bands and are determined by minimizing a norm of the difference between snapshot spectra and their averages over the energy elements. We begin by presenting the theory of the FEDS method. We then compare to continuous-energy Monte Carlo for one-dimensional slab and two-dimensional pin-cell problem. We find FEDS to be accurate and efficient at producing quantities of interest such as reaction rates and eigenvalues. Results show that FEDS converges at a rate that is approximately first-order in the number of energy elements and that FEDS is less sensitive to weighting spectrum than standard MG.

  20. Carbon-Neutral Energy Supply and Energy Demand-Reduction Technology Needed for Continued Economic Growth Without Dangerous Interference in the Climate System

    Science.gov (United States)

    Hoffert, M. I.; Caldeira, K.

    2007-12-01

    Stabilization of atmospheric CO2 at levels likely to avoid unacceptable climate risk will require a major transformation in the ways we produce and use energy. Most of our energy will need to come from sources that do not emit carbon dioxide to the atmosphere and that energy will need to be used efficiently. The required reduction of carbon dioxide emissions as global energy consumption and GDP grow imposes quantitative requirements on some combination of carbon-neutral primary power and energy demand reduction. (Emission reductions are expressed relative to an implicit or explicit baseline; explicit being better for policy-making. Energy demand reduction involves both efficiency improvements and lifestyle changes.) These requirements can be expressed as CO2 emission reductions needed, or as carbon-neutral primary power production needed combined with power not used by virtue of increased energy end use efficiency or lifestyle changes ("negawatts"), always subject to some reasonably well-characterized uncertainty limits. Climatic changes thus far have been closer to the more extreme zone of the climatic uncertainty envelope of global warming indicating the potential for disastrous impacts by mid-century and beyond for business-as-usual. Emission reductions needed to avoid "dangerous interference in the climate system" imply a revolutionary change in the global energy system beginning now; particularly ominous are massive conventional coal-fired electric power energy infrastructures under construction by the US, China & India. Strong arguments, based on physical science considerations, exist for prompt measures such as (1) an immediate moratorium on coal-fired plants that don't sequester CO2, (2) a gradually increasing price on carbon emissions and (3) regulatory standards, for example, that would encourage utilities and car manufacturers to improve efficiency, and (4) Apollo-scale R & D projects beginning now to develop sustainable carbon-neutral power that can be

  1. In vivo continuous and simultaneous monitoring of brain energy substrates with a multiplex amperometric enzyme-based biosensor device

    NARCIS (Netherlands)

    De Lima Braga Lopes Cordeiro, Carlos; de Vries, M.G.; Ngabi, W; Oomen, P.E.; Cremers, T.I.F.H.; Westerink, B.H.C.

    2015-01-01

    Enzyme-based amperometric biosensors are widely used for monitoring key biomarkers. In experimental neuroscience there is a growing interest in in vivo continuous and simultaneous monitoring of metabolism-related biomarkers, like glucose, lactate and pyruvate. The use of multiplex biosensors will pr

  2. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  3. Note: Proton microbeam formation with continuously variable kinetic energy using a compact system for three-dimensional proton beam writing

    Energy Technology Data Exchange (ETDEWEB)

    Ohkubo, T., E-mail: ohkubo.takeru@jaea.go.jp; Ishii, Y. [Department of Advanced Radiation Technology, Takasaki Advanced Radiation Research Institute, Japan Atomic Energy Agency, 1233 Watanuki-machi, Takasaki, Gunma 370-1292 (Japan)

    2015-03-15

    A compact focused gaseous ion beam system has been developed to form proton microbeams of a few hundreds of keV with a penetration depth of micrometer range in 3-dimensional proton beam writing. Proton microbeams with kinetic energies of 100-140 keV were experimentally formed on the same point at a constant ratio of the kinetic energy of the object side to that of the image side. The experimental results indicate that the beam diameters were measured to be almost constant at approximately 6 μm at the same point with the kinetic energy range. These characteristics of the system were experimentally and numerically demonstrated to be maintained as long as the ratio was constant.

  4. Study of the triton-burnup process in different JET scenarios using neutron monitor based on CVD diamond

    Science.gov (United States)

    Nemtsev, G.; Amosov, V.; Meshchaninov, S.; Popovichev, S.; Rodionov, R.

    2016-11-01

    We present the results of analysis of triton burn-up process using the data from diamond detector. Neutron monitor based on CVD diamond was installed in JET torus hall close to the plasma center. We measure the part of 14 MeV neutrons in scenarios where plasma current varies in a range of 1-3 MA. In this experiment diamond neutron monitor was also able to detect strong gamma bursts produced by runaway electrons arising during the disruptions. We can conclude that CVD diamond detector will contribute to the study of fast particles confinement and help predict the disruption events in future tokamaks.

  5. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  6. Performance analysis of a continuous serpentine flow reactor for electrochemical oxidation of synthetic and real textile wastewater: Energy consumption, mass transfer coefficient and economic analysis.

    Science.gov (United States)

    Pillai, Indu M Sasidharan; Gupta, Ashok K

    2017-05-15

    A continuous flow electrochemical reactor was developed, and its application was tested for the treatment of textile wastewater. A parallel plate configuration with serpentine flow was chosen for the continuous flow reactor. Uniparameter optimization was carried out for electrochemical oxidation of synthetic and real textile wastewater (collected from the inlet of the effluent treatment plant). Chemical Oxygen Demand (COD) removal efficiency of 90% was achieved for synthetic textile wastewater (initial COD - 780 mg L(-1)) at a flow rate of 500 mL h(-1) (retention time of 6 h) and a current density of 1.15 mA cm(-2) and the energy consumption for the degradation was 9.2 kWh (kg COD)(-1). The complete degradation of real textile wastewater (initial COD of 368 mg L(-1)) was obtained at a current density of 1.15 mA cm(-2), NaCl concentration of 1 g L(-1) and retention time of 6 h. Energy consumption and mass transfer coefficient of the reactions were calculated. The continuous flow reactor performed better than batch reactor with reference to energy consumption and economy. The overall treatment cost for complete COD removal of real textile wastewater was 5.83 USD m(-3).

  7. Grid connected integrated community energy system. Phase II: final stage 2 report. Outline specifications of cogeneration plant; continued

    Energy Technology Data Exchange (ETDEWEB)

    1978-03-22

    Specifications are presented for the electrical equipment, site preparation, building construction and mechanical systems for a dual-purpose power plant to be located on the University of Minnesota campus. This power plant will supply steam and electrical power to a grid-connected Integrated Community Energy System. (LCL)

  8. Low intensity, continuous wave photodoping of ZnO quantum dots - photon energy and particle size effects.

    Science.gov (United States)

    Aguirre, Matías E; Municoy, S; Grela, M A; Colussi, A J

    2017-02-08

    The unique properties of semiconductor quantum dots (QDs) have found application in the conversion of solar to chemical energy. How the relative rates of the redox processes that control QD photon efficiencies depend on the particle radius (r) and photon energy (Eλ), however, is not fully understood. Here, we address these issues and report the quantum yields (Φs) of interfacial charge transfer and electron doping in ZnO QDs capped with ethylene glycol (EG) as a function of r and Eλ in the presence and absence of methyl viologen (MV(2+)) as an electron acceptor, respectively. We found that Φs for the oxidation of EG are independent of Eλ and photon fluence (φλ), but markedly increase with r. The independence of Φs on φλ ensures that QDs are never populated by more than one electron-hole pair, thereby excluding Auger-type terminations. We show that these findings are consistent with the operation of an interfacial redox process that involves thermalized carriers in the Marcus inverted region. In the absence of MV(2+), QDs accumulate electrons up to limiting volumetric densities ρe,∞ that depend sigmoidally on excess photon energy E* = Eλ - EBG(r), where EBG(r) is the r-dependent bandgap energy. The maximum electron densities: ρev,∞ ∼ 4 × 10(20) cm(-3), are reached at E* > 0.5 eV, independent of the particle radius.

  9. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Science.gov (United States)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  10. Efficient and Accurate Calculation of Burnup Problems with Short-Lived Nuclides by a Krylov Subspace Method with the Newton Divided Difference

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hee; Cho, Nam Zin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Nowadays lattice physics codes tend to utilize a detailed burnup chain including short-lived nuclides in order to perform more accurate burnup calculations. But, since production codes, for example, ORIGEN2, take account of nuclides which have relatively long half-life, it is inappropriate for such detailed burnup chain calculation. To enhance that drawback, many matrix exponential calculation methods have been developed. Recently, a Krylov subspace method with the PADE approximation was used. In this paper, a Krylov subspace method based on spectral decomposition property of the matrix function theory with the Newton divided difference (NDD) is introduced. It is tested with a sample problem and compared with simple Taylor expansion method

  11. Chemical states of fission products in irradiated (U 0.3Pu 0.7)C 1+ x fuel at high burn-ups

    Science.gov (United States)

    Agarwal, Renu; Venugopal, V.

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel, (U 0.3Pu 0.7)C 1+ x, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  12. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    Science.gov (United States)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. SUP>235U-rich agglomerates, SUP>235U-poor areas, an intermediate phase with intermediate 235U concentrations. Short fuel rods were fabricated with these pellets. The main characteristics of these fuel rods are shown in Table 1.These rods were irradiated to high burn-ups in the IFA-609/626 of the HBWR and then one was irradiated in the IFA-702 for 100 days. Fig. 2 shows the irradiation history of this fuel. The final average burn-up of the rod was 69 GWd/tU. Due to the flux differences along the rod, however, the average burn-up of the cross section examined was 63 GWd/tU. This fuel

  13. The Time-Frequency Energy Attenuation Factor and Its Application on the Basis of Gauss Linear Frequency-Modulated Continuous Wavelet Transform

    Institute of Scientific and Technical Information of China (English)

    Liu Xiqiang; Shen Ping; Li Hong; Shan Changlun; Ji Aidong; Zhang Ping; Cai Mingjun

    2004-01-01

    Based on the Gauss linear frequency-modulated wavelet transform, a new characteristic index is presented, namely time-frequency energy attenuation factor which can reflect the difference features of waveform in earthquake focus mechanism, wave traveling path and its attenuation characteristics in focal area or near field. In order to test its validity, we select the natural earthquakes and explosion or collapse events whose focus mechanisms vary obviously, and some natural earthquakes located at the same site or in a very small area. The study indicates that the time-frequency energy attenuation factors of the natural earthquakes are obviously different with that of explosion or collapse events, and the change of the time-frequency energy attenuation factors is relatively stable for the earthquakes under the normal seismicity background. Using the above-mentioned method, it is expected to offer a useful criterion for strong earthquake prediction by continuous earthquake observation.

  14. Results of the studies on energy deposition in IR6 superconducting magnets from continuous beam loss on the TCDQ system

    CERN Document Server

    Bracco, C; Presland, A; Redaelli, S; Sarchiapone, L; Weiler, T

    2007-01-01

    A single sided mobile graphite diluter block TCDQ, in combination with a two-sided secondary collimator TCS and an iron shield TCDQM, will be installed in front of the superconducting quadrupole Q4 magnets in IR6, in order to protect it and other downstream LHC machine elements from destruction in the event of a beam dump that is not synchronised with the abort gap. The TCDQ will be positioned close to the beam, and will intercept the particles from the secondary halo during low beam lifetime. Previous studies (1-4) have shown that the energy deposited in the Q4 magnet coils can be close to or above the quench limit. In this note the results of the latest FLUKA energy deposition simulations for Beam 2 are described, including an upgrade possibility for the TCDQ system with an additional shielding device. The results are discussed in the context of the expected performance levels for the different phases of LHC operation.

  15. Signature of a continuous quantum phase transition in non-equilibrium energy absorption: Footprints of criticality on higher excited states.

    Science.gov (United States)

    Bhattacharyya, Sirshendu; Dasgupta, Subinay; Das, Arnab

    2015-11-16

    Understanding phase transitions in quantum matters constitutes a significant part of present day condensed matter physics. Quantum phase transitions concern ground state properties of many-body systems, and hence their signatures are expected to be pronounced in low-energy states. Here we report signature of a quantum critical point manifested in strongly out-of-equilibrium states with finite energy density with respect to the ground state and extensive (subsystem) entanglement entropy, generated by an external pulse. These non-equilibrium states are evidently completely disordered (e.g., paramagnetic in case of a magnetic ordering transition). The pulse is applied by switching a coupling of the Hamiltonian from an initial value (λI) to a final value (λF) for sufficiently long time and back again. The signature appears as non-analyticities (kinks) in the energy absorbed by the system from the pulse as a function of λF at critical-points (i.e., at values of λF corresponding to static critical-points of the system). As one excites higher and higher eigenstates of the final Hamiltonian H(λF) by increasing the pulse height (|λF - λI|), the non-analyticity grows stronger monotonically with it. This implies adding contributions from higher eigenstates help magnifying the non-analyticity, indicating strong imprint of the critical-point on them. Our findings are grounded on exact analytical results derived for Ising and XY chains in transverse field.

  16. Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-01

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

  18. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    Science.gov (United States)

    Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

    2017-09-01

    Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  19. Analysing drying unit performance in a continuous pharmaceutical manufacturing line by means of mass – Energy balances

    DEFF Research Database (Denmark)

    Mortier, Séverine Thérèse F.C.; Gernaey, Krist; De Beer, Thomas De Beer

    2014-01-01

    used. In this paper the data of the six-segmented fluidized bed dryer in the line are used for the development and evaluation of a mass and energy balance. The objectives are multiple: (1) prediction of the moisture content of the granules leaving the dryer solely based on the currently logged data...... in the fluidized bed can be used to predict the gas temperature in different horizontal sections of the dryer. An extra sensor measuring the gas temperature and the humidity at the wet transfer line would increase the accuracy of the calculations. An extra gas velocity sensor at the outlet would be useful...

  20. Monopolies, liberalization, energy turnaround. (Dis)continuities in the electricity market design; Monopole, Liberalisierung, Energiewende. (Dis-)Kontinuitaeten im Strommarktdesign

    Energy Technology Data Exchange (ETDEWEB)

    Grashof, Katherina; Zipp, Alexander [Institut fuer ZukunftsEnergieSysteme (IZES), Saarbruecken (Germany); Jachmann, Henning [Zentrum fuer Sonnenenergie- und Wasserstoff-Forschung Baden-Wuerttemberg (ZSW), Stuttgart (Germany); Wille-Haussmann, Bernhard [Fraunhofer-Institut fuer Solare Energiesysteme (ISE), Freiburg im Breisgau (Germany); Lechtenboehmer, Stefan [Wuppertal Institut fuer Klima, Umwelt, Energie GmbH, Wuppertal (Germany)

    2015-04-15

    After a long period of stability, the electricity industry is in the past 15 years, in a major state of flux. First, the switching of state-monitored and regulated regional monopolies to liberalized producer and consumer markets. At the moment we are in a similar change from conventional to renewable energy production. Below the main question will be addressed whether the paradigms of the individual phases are compatible, which still have their place and which should be modified. Moreover, it is shown that the current market design of the future cannot be designed on a blank sheet, but existing structures have to be considered. Given the stage of monopolies, the liberalization and the started energy turnaround respectively in terms of their sector structure, dominant generation technologies, the interaction between production and load and characteristic elements of market design and regulation are presented. Subsequently, a preliminary answer is given to the question raised. [German] Nach einer langen Phase der Stabilitaet ist die Stromwirtschaft in den vergangenen 15 Jahren stark in Bewegung geraten. Zunaechst stand der Wechsel von staatlich ueberwachten und regulierten Gebietsmonopolen hin zu liberalisierten Erzeuger und Verbrauchermaerkten an. Im Moment befinden wir uns in einem aehnlichen Umbruch, weg von konventioneller hin zu erneuerbarer Energieerzeugung. Im Folgenden soll der Leitfrage nachgegangen werden, ob die Paradigmen der einzelnen Phasen miteinander vereinbar sind, welche noch immer ihre Daseinsberechtigung haben und welche modifiziert werden sollten. Darueber hinaus wird gezeigt, dass das Strommarktdesign der Zukunft nicht auf einem leeren Blatt entworfen werden kann, sondern bestehende Strukturen zu beruecksichtigen sind. Dazu werden die Phase der Monopolwirtschaft, der Liberalisierung sowie der begonnenen Energiewende jeweils hinsichtlich ihrer Sektor Struktur, dominierenden Erzeugungstechnologien, des Zusammenspiels zwischen Erzeugung und Last

  1. Microstructure evolution of nanostructured and submicrometric porous refractory ceramics induced by a continuous high-energy proton beam

    Science.gov (United States)

    Fernandes, Sandrina; Bruetsch, Roland; Catherall, Richard; Groeschel, Friedrich; Guenther-Leopold, Ines; Lettry, Jacques; Manfrin, Enzo; Marzari, Stefano; Noah, Etam; Sgobba, Stefano; Stora, Thierry; Zanini, Luca

    2011-09-01

    The production of radioactive ion beams by the isotope mass separation online (ISOL) method requires a fast diffusion and effusion of nuclear products from thick refractory target materials under high-energy particle beam irradiation. A new generation of ISOL nanostructured and submicrometric porous materials have been developed, exhibiting enhanced release of exotic isotopes, compared to previously used conventional micrometric materials. A programme was developed at PSI within the framework sof the Design Study of EURISOL, the next generation European ISOL-type facility to study aging under irradiation on porous ceramic pellets and dense thin metal foils at high temperatures. Ceramic oxides and carbide samples underwent proton damage with fluence up to 3.0 × 10 20 and 1.3 × 10 21 cm -2 respectively. The post-irradiation examination on Al 2O 3, Y 2O 3 and SiC - C nanotube composite matrices show a proton-induced densification region in which a moderate grain growth occurred. The microstructural evolution effects were associated to the combination of radiation-enhanced diffusion and thermal diffusion. The irradiated Al 2O 3 shows higher sintering rates than in similar non-irradiation isothermal conditions, in particular at the lowest irradiation temperature, subjected to a proton fluence inferior to 1.1 × 10 15 cm -2. The apparent activation energy for its sintering controlling mechanism was found to be between 44 and 88 kJ mol -1. However, despite the enhanced sintering, shrinkage and increased grain growth, the selected nanostructured and submicrometric TARPIPE materials did not display an average grain diameter above 2 μm, which confirms that these materials are suited as production targets for present and next generation ISOL facilities.

  2. Modelling and Validation of Synthesis of Poly Lactic Acid Using an Alternative Energy Source through a Continuous Reactive Extrusion Process

    Directory of Open Access Journals (Sweden)

    Satya P. Dubey

    2016-04-01

    Full Text Available PLA is one of the most promising bio-compostable and bio-degradable thermoplastic polymers made from renewable sources. PLA is generally produced by ring opening polymerization (ROP of lactide using the metallic/bimetallic catalyst (Sn, Zn, and Al or other organic catalysts in a suitable solvent. In this work, reactive extrusion experiments using stannous octoate Sn(Oct2 and tri-phenyl phosphine (PPh3 were considered to perform ROP of lactide. Ultrasound energy source was used for activating and/or boosting the polymerization as an alternative energy (AE source. Ludovic® software, designed for simulation of the extrusion process, had to be modified in order to simulate the reactive extrusion of lactide and for the application of an AE source in an extruder. A mathematical model for the ROP of lactide reaction was developed to estimate the kinetics of the polymerization process. The isothermal curves generated through this model were then used by Ludovic software to simulate the “reactive” extrusion process of ROP of lactide. Results from the experiments and simulations were compared to validate the simulation methodology. It was observed that the application of an AE source boosts the polymerization of lactide monomers. However, it was also observed that the predicted residence time was shorter than the experimental one. There is potentially a case for reducing the residence time distribution (RTD in Ludovic® due to the ‘liquid’ monomer flow in the extruder. Although this change in parameters resulted in validation of the simulation, it was concluded that further research is needed to validate this assumption.

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

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

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

  4. Long-term bio-H2 and bio-CH4 production from food waste in a continuous two-stage system: Energy efficiency and conversion pathways.

    Science.gov (United States)

    Algapani, Dalal E; Qiao, Wei; di Pumpo, Francesca; Bianchi, David; Wandera, Simon M; Adani, Fabrizio; Dong, Renjie

    2017-05-29

    Anaerobic digestion is a well-established technology for treating organic waste, but it is still under challenge for food waste due to process stability problems. In this work, continuous H2 and CH4 production from canteen food waste (FW) in a two-stage system were successfully established by optimizing process parameters. The optimal hydraulic retention time was 5d for H2 and 15d for CH4. Overall, around 59% of the total COD in FW was converted into H2 (4%) and into CH4 (55%). The fluctuations of FW characteristics did not significantly affect process performance. From the energy point view, the H2 reactor contributed much less than the methane reactor to total energy balance, but it played a key role in maintaining the stability of anaerobic treatment of food waste. Microbial characterization indicated that methane formation was through syntrophic acetate oxidation combined with hydrogenotrophic methanogenesis pathway. Copyright © 2017. Published by Elsevier Ltd.

  5. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

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

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

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

  7. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Protsik, R.; Lewellen, J.W. (eds.)

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the techniques for modeling the three-dimensional benchmark geometry, and sensitivity studies were carried out to determine the performance parameter sensitivities to changes in the neutronics and burnup specifications. The results of the Benchmark Four calculations indicated that a linked RZ-XY (Hex) two-dimensional representation of the benchmark model geometry can be used to predict mass balance data, power distributions, regionwise fuel exposure data and burnup reactivities with good accuracy when compared with the results of direct three-dimensional computations. Most of the small differences in the results of the benchmark analyses by the different participants were attributed to ambiguities in carrying out the regionwise flux renormalization calculations throughout the burnup step.

  8. 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.

  9. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10{sup 25} m{sup -2}, respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

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

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

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

  11. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  12. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  13. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [EDF/SEPTEN Villeurbanne (France); GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  14. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  15. An Integrated Optimal Energy Management/Gear-Shifting Strategy for an Electric Continuously Variable Transmission Hybrid Powertrain Using Bacterial Foraging Algorithm

    Directory of Open Access Journals (Sweden)

    Syuan-Yi Chen

    2016-01-01

    Full Text Available This study developed an integrated energy management/gear-shifting strategy by using a bacterial foraging algorithm (BFA in an engine/motor hybrid powertrain with electric continuously variable transmission. A control-oriented vehicle model was constructed on the Matlab/Simulink platform for further integration with developed control strategies. A baseline control strategy with four modes was developed for comparison with the proposed BFA. The BFA was used with five bacterial populations to search for the optimal gear ratio and power-split ratio for minimizing the cost: the equivalent fuel consumption. Three main procedures were followed: chemotaxis, reproduction, and elimination-dispersal. After the vehicle model was integrated with the vehicle control unit with the BFA, two driving patterns, the New European Driving Cycle and the Federal Test Procedure, were used to evaluate the energy consumption improvement and equivalent fuel consumption compared with the baseline. The results show that [18.35%,21.77%] and [8.76%,13.81%] were improved for the optimal energy management and integrated optimization at the first and second driving cycles, respectively. Real-time platform designs and vehicle integration for a dynamometer test will be investigated in the future.

  16. Energy

    CERN Document Server

    Foland, Andrew Dean

    2007-01-01

    Energy is the central concept of physics. Unable to be created or destroyed but transformable from one form to another, energy ultimately determines what is and isn''t possible in our universe. This book gives readers an appreciation for the limits of energy and the quantities of energy in the world around them. This fascinating book explores the major forms of energy: kinetic, potential, electrical, chemical, thermal, and nuclear.

  17. DFI oxyfuel process for saving energy and improving the performance and quality of continuous strip lines; DFI-Oxyfuel-Verfahren zur Energieeinsparung, Leistungs- und Qualitaetssteigerung von Banddurchlaufanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Eichelkraut, H. [ThyssenKrupp Steel AG, Duisburg (Germany). Standort Bruckhausen; Heiler, H.J. [ThyssenKrupp Steel AG, Finnentrop (Germany); Domels, H.P.; Hoegner, W. [ThyssenKrupp Steel AG, Duisburg (Germany). Energie/Anlagenwirtschaft

    2007-07-01

    The development of the 'Direct Flame Impingement (DFI)-Oxyfuel' process - in which an oxyfuel (oxygen-fuel) flame impinges directly on the material to be heated - represents a further development of furnace technology for continuous strip lines. With assistance from the cooperation partner Linde, the process was used on a hot-dip galvanizing line for the first time at ThyssenKrupp Steel's Finnentrop plant. Right from the start it produced outstanding results in terms of increased throughput, product quality, plant quality and energy efficiency and thus also a reduction in direct CO{sub 2} emissions. In the meantime, this technology also is being used at an additional strip galvanizing and aluminizing facility in the Duisburg-Bruckhausen plant. (orig.)

  18. Effects of intermittent compared to continuous energy restriction on short-term weight loss and long-term weight loss maintenance.

    Science.gov (United States)

    Keogh, J B; Pedersen, E; Petersen, K S; Clifton, P M

    2014-06-01

    Effective strategies are needed to help individuals lose weight and maintain weight loss. The primary aim of this study was to investigate the effect of intermittent energy restriction (IER) compared to continuous energy restriction (CER) on weight loss after 8 weeks and weight loss maintenance after 12 months. Secondary aims were to determine changes in waist and hip measurements and diet quality. In a randomized parallel study, overweight and obese (body mass index [BMI] ≥ 27 kg m(-2)) women were stratified by age and BMI before randomization. Participants undertook an 8-week intensive period with weight, waist and hip circumference measured every 2 weeks, followed by 44 weeks of independent dieting. A food frequency questionnaire was completed at baseline and 12 months, from which diet quality was determined. Weight loss was not significantly different between the two groups at 8 weeks (-3.2 ± 2.1 kg CER, n = 20, -2.0 ± 1.9 kg IER, n = 25; P = 0.06) or at 12 months (-4.2 ± 5.6 kg CER, n = 17 -2.1 ± 3.8 kg IER, n = 19; P = 0.19). Weight loss between 8 and 52 weeks was -0.7 ± 49 kg CER vs. -1 ± 1.1 kg IER; P = 0.6. Waist and hip circumference decreased significantly with time (P intermittent dieting was as effective as continuous dieting over 8 weeks and for weight loss maintenance at 12 months. This may be useful for individuals who find CER too difficult to maintain.

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

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

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

  20. Ubiquitous Transgenic Overexpression of C-C Chemokine Ligand 2: A Model to Assess the Combined Effect of High Energy Intake and Continuous Low-Grade Inflammation

    Science.gov (United States)

    Rodríguez-Gallego, Esther; Hernández-Aguilera, Anna; Mariné-Casadó, Roger; Rull, Anna; Beltrán-Debón, Raúl; Menendez, Javier A.; Vazquez-Martin, Alejandro; Sirvent, Juan J.; Martín-Paredero, Vicente; Corbí, Angel L.; Sierra-Filardi, Elena; Aragonès, Gerard; García-Heredia, Anabel; Camps, Jordi; Alonso-Villaverde, Carlos; Joven, Jorge

    2013-01-01

    Excessive energy management leads to low-grade, chronic inflammation, which is a significant factor predicting noncommunicable diseases. In turn, inflammation, oxidation, and metabolism are associated with the course of these diseases; mitochondrial dysfunction seems to be at the crossroads of mutual relationships. The migration of immune cells during inflammation is governed by the interaction between chemokines and chemokine receptors. Chemokines, especially C-C-chemokine ligand 2 (CCL2), have a variety of additional functions that are involved in the maintenance of normal metabolism. It is our hypothesis that a ubiquitous and continuous secretion of CCL2 may represent an animal model of low-grade chronic inflammation that, in the presence of an energy surplus, could help to ascertain the afore-mentioned relationships and/or to search for specific therapeutic approaches. Here, we present preliminary data on a mouse model created by using targeted gene knock-in technology to integrate an additional copy of the CCl2 gene in the Gt(ROSA)26Sor locus of the mouse genome via homologous recombination in embryonic stem cells. Short-term dietary manipulations were assessed and the findings include metabolic disturbances, premature death, and the manipulation of macrophage plasticity and autophagy. These results raise a number of mechanistic questions for future study. PMID:24453432

  1. Ubiquitous transgenic overexpression of C-C chemokine ligand 2: a model to assess the combined effect of high energy intake and continuous low-grade inflammation.

    Science.gov (United States)

    Rodríguez-Gallego, Esther; Riera-Borrull, Marta; Hernández-Aguilera, Anna; Mariné-Casadó, Roger; Rull, Anna; Beltrán-Debón, Raúl; Luciano-Mateo, Fedra; Menendez, Javier A; Vazquez-Martin, Alejandro; Sirvent, Juan J; Martín-Paredero, Vicente; Corbí, Angel L; Sierra-Filardi, Elena; Aragonès, Gerard; García-Heredia, Anabel; Camps, Jordi; Alonso-Villaverde, Carlos; Joven, Jorge

    2013-01-01

    Excessive energy management leads to low-grade, chronic inflammation, which is a significant factor predicting noncommunicable diseases. In turn, inflammation, oxidation, and metabolism are associated with the course of these diseases; mitochondrial dysfunction seems to be at the crossroads of mutual relationships. The migration of immune cells during inflammation is governed by the interaction between chemokines and chemokine receptors. Chemokines, especially C-C-chemokine ligand 2 (CCL2), have a variety of additional functions that are involved in the maintenance of normal metabolism. It is our hypothesis that a ubiquitous and continuous secretion of CCL2 may represent an animal model of low-grade chronic inflammation that, in the presence of an energy surplus, could help to ascertain the afore-mentioned relationships and/or to search for specific therapeutic approaches. Here, we present preliminary data on a mouse model created by using targeted gene knock-in technology to integrate an additional copy of the CCl2 gene in the Gt(ROSA)26Sor locus of the mouse genome via homologous recombination in embryonic stem cells. Short-term dietary manipulations were assessed and the findings include metabolic disturbances, premature death, and the manipulation of macrophage plasticity and autophagy. These results raise a number of mechanistic questions for future study.

  2. Ubiquitous Transgenic Overexpression of C-C Chemokine Ligand 2: A Model to Assess the Combined Effect of High Energy Intake and Continuous Low-Grade Inflammation

    Directory of Open Access Journals (Sweden)

    Esther Rodríguez-Gallego

    2013-01-01

    Full Text Available Excessive energy management leads to low-grade, chronic inflammation, which is a significant factor predicting noncommunicable diseases. In turn, inflammation, oxidation, and metabolism are associated with the course of these diseases; mitochondrial dysfunction seems to be at the crossroads of mutual relationships. The migration of immune cells during inflammation is governed by the interaction between chemokines and chemokine receptors. Chemokines, especially C-C-chemokine ligand 2 (CCL2, have a variety of additional functions that are involved in the maintenance of normal metabolism. It is our hypothesis that a ubiquitous and continuous secretion of CCL2 may represent an animal model of low-grade chronic inflammation that, in the presence of an energy surplus, could help to ascertain the afore-mentioned relationships and/or to search for specific therapeutic approaches. Here, we present preliminary data on a mouse model created by using targeted gene knock-in technology to integrate an additional copy of the CCl2 gene in the Gt(ROSA26Sor locus of the mouse genome via homologous recombination in embryonic stem cells. Short-term dietary manipulations were assessed and the findings include metabolic disturbances, premature death, and the manipulation of macrophage plasticity and autophagy. These results raise a number of mechanistic questions for future study.

  3. Development and implementation of a pilot network for continuous monitoring of electric energy; Desarrollo e implantacion de una red piloto de monitoreo continuo de energia electrica

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez Cifuentes, A.; Cortes Eslava, A. [Facultad de Ingenieria, UNAM, Mexico, D. F. (Mexico)

    1997-12-31

    The ``Programa Universitario de Energia (PUE)`` (University Energy Program) of the Universidad Nacional Autonoma de Mexico (UNAM) through the project ``Ecologic Control of The University Campus`` has the objective of instrumenting policies for the rational use of the electric energy in the UNAM installations. The PUE has performed a series of activities aimed at the establishment of an integral program for the electric energy management in the UNAM. This program has to consider the measures for energy saving (MES) of general kind and to promote a series of specific measures to be adequated and implemented for each one of the departments. The success in establishing a series of MES compels to quantify its effects, to register and to analyze them to be in position of enhancing them, since it must be kept in mind that something that can not be measured can not be improved. For this purpose the PUE is developing a continuous monitoring pilot network of electric energy with local and remote access and an automated metering system. Two monitoring schemes are experimented, one commercial metering equipment and another one with own metering technology. In this paper the characteristics of the proposed monitoring schemes are described. [Espanol] El Programa Universitario de Energia (PUE) de la Universidad Nacional Autonoma de Mexico (UNAM), a traves del proyecto {sup C}ontrol Ecologico del Campus Universitario{sup ,} tiene como objetivo instrumentar politicas de uso racional de la energia en las instalaciones universitarias. El PUE ha realizado una serie de actividades encaminadas a establecer un programa integral de administracion de la energia electrica en la UNAM. El programa mencionado debe considerar medidas de ahorro de energia (MAE) de tipo general y promover una serie de medidas particulares, para ser adecuadas e implantadas por cada una de las dependencias. El exito en el establecimiento de una serie de MAE, obliga a cuantificar sus efectos, registrarse y

  4. Continuous auditing & continuous monitoring : Continuous value?

    NARCIS (Netherlands)

    van Hillo, Rutger; Weigand, Hans; Espana, S; Ralyte, J; Souveyet, C

    2016-01-01

    Advancements in information technology, new laws and regulations and rapidly changing business conditions have led to a need for more timely and ongoing assurance with effectively working controls. Continuous Auditing (CA) and Continuous Monitoring (CM) technologies have made this possible by obtain

  5. Continuous auditing & continuous monitoring : Continuous value?

    NARCIS (Netherlands)

    van Hillo, Rutger; Weigand, Hans; Espana, S; Ralyte, J; Souveyet, C

    2016-01-01

    Advancements in information technology, new laws and regulations and rapidly changing business conditions have led to a need for more timely and ongoing assurance with effectively working controls. Continuous Auditing (CA) and Continuous Monitoring (CM) technologies have made this possible by obtain

  6. Continuous auditing & continuous monitoring : Continuous value?

    NARCIS (Netherlands)

    van Hillo, Rutger; Weigand, Hans; Espana, S; Ralyte, J; Souveyet, C

    2016-01-01

    Advancements in information technology, new laws and regulations and rapidly changing business conditions have led to a need for more timely and ongoing assurance with effectively working controls. Continuous Auditing (CA) and Continuous Monitoring (CM) technologies have made this possible by

  7. Energy

    CERN Document Server

    Robertson, William C

    2002-01-01

    Confounded by kinetic energy? Suspect that teaching about simple machines isn t really so simple? Exasperated by electricity? If you fear the study of energy is beyond you, this entertaining book will do more than introduce you to the topic. It will help you actually understand it. At the book s heart are easy-to-grasp explanations of energy basics work, kinetic energy, potential energy, and the transformation of energy and energy as it relates to simple machines, heat energy, temperature, and heat transfer. Irreverent author Bill Robertson suggests activities that bring the basic concepts of energy to life with common household objects. Each chapter ends with a summary and an applications section that uses practical examples such as roller coasters and home heating systems to explain energy transformations and convection cells. The final chapter brings together key concepts in an easy-to-grasp explanation of how electricity is generated. Energy is the second book in the Stop Faking It! series published by NS...

  8. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  9. Determination of burnup grade of fuel plates by gamma spectrometry; Determinacao do grau de queima em elementos combustiveis tipo placa por meio de espectrometria gama

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

    This work describes absolute burnup measurements on spent MTR fuel elements by means of non-destructive gamma-ray spectroscopy which correlates activities of radioactive fission products with the fissioned mass of {sup 235} U. Experiments based on such method were performed at the storage pool area of the IEA-R1 research reactor. The obtained results were compared with calculational ones based on neutronics. (author) 7 refs., 6 figs., 1 tab.; e-mail: laaterre at net.ipen.br

  10. The physics of transverse mode instability-induced nonlinear phase distortions in large area optical fiber amplifiers and their mitigation with applications in scaling of pulsed and continuous wave high-energy lasers

    Science.gov (United States)

    2016-12-13

    their mitigation with applications in scaling of pulsed and continuous- wave high- energy lasers Balaji Srinivasan INDIAN INSTITUTE OF TECHNOLOGY...high- energy lasers 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA2386-15-1-5044 5c.  PROGRAM ELEMENT NUMBER 61102F 6. AUTHOR(S) Balaji Srinivasan 5d...use of vortex beams to mitigate thermal mode instability in high energy fiber amplifiers. The investigation is carried out through (1) the

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

    Science.gov (United States)

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

    2015-03-01

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

  12. Measurement of the composition of noble-metal particles in high-burnup CANDU fuel by wavelength dispersive X-ray microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Szostak, F.J

    1999-09-01

    An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely {sup 235}U fission to mainly {sup 239}Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)

  13. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-03-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  14. Very High Fuel Economy, Heavy Duty, Constant Speed, Truck Engine Optimized Via Unique Energy Recovery Turbines and Facilitated High Efficiency Continuously Variable Drivetrain

    Energy Technology Data Exchange (ETDEWEB)

    Bahman Habibzadeh

    2010-01-31

    The project began under a corporative agreement between Mack Trucks, Inc and the Department of Energy starting from September 1, 2005. The major objective of the four year project is to demonstrate a 10% efficiency gain by operating a Volvo 13 Litre heavy-duty diesel engine at a constant or narrow speed and coupled to a continuously variable transmission. The simulation work on the Constant Speed Engine started on October 1st. The initial simulations are aimed to give a basic engine model for the VTEC vehicle simulations. Compressor and turbine maps are based upon existing maps and/or qualified, realistic estimations. The reference engine is a MD 13 US07 475 Hp. Phase I was completed in May 2006 which determined that an increase in fuel efficiency for the engine of 10.5% over the OICA cycle, and 8.2% over a road cycle was possible. The net increase in fuel efficiency would be 5% when coupled to a CVT and operated over simulated highway conditions. In Phase II an economic analysis was performed on the engine with turbocompound (TC) and a Continuously Variable Transmission (CVT). The system was analyzed to determine the payback time needed for the added cost of the TC and CVT system. The analysis was performed by considering two different production scenarios of 10,000 and 60,000 units annually. The cost estimate includes the turbocharger, the turbocompound unit, the interstage duct diffuser and installation details, the modifications necessary on the engine and the CVT. Even with the cheapest fuel and the lowest improvement, the pay back time is only slightly more than 12 months. A gear train is necessary between the engine crankshaft and turbocompound unit. This is considered to be relatively straight forward with no design problems.

  15. Continuous-Energy Adjoint Flux and Perturbation Calculation using the Iterated Fission Probability Method in Monte Carlo Code TRIPOLI-4® and Underlying Applications

    Science.gov (United States)

    Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.; Malvagi, F.

    2014-06-01

    Pile-oscillation experiments are performed in the MINERVE reactor at the CEA Cadarache to improve nuclear data accuracy. In order to precisely calculate small reactivity variations (experiments, a reference calculation need to be achieved. This calculation may be accomplished using the continuous-energy Monte Carlo code TRIPOLI-4® by using the eigenvalue difference method. This "direct" method has shown limitations in the evaluation of very small reactivity effects because it needs to reach a very small variance associated to the reactivity in both states. To answer this problem, it has been decided to implement the exact perturbation theory in TRIPOLI-4® and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4® is described. To illustrate the effciency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the "direct" estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the "direct" method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. Other applications of

  16. Continuous-time signals

    CERN Document Server

    Shmaliy, Yuriy

    2006-01-01

    Gives a modern description of continuous-time deterministic signals Signal formation techniquesTime vs. frequency and frequency vs. time analysisCorrelation and energy analysisNarrowband signals and sampling.

  17. Influence of Continuous Flow Microwave Pre-Treatment on Anaerobic Digestion of Secondary Thickened Sludge for Sustainable Energy Recovery in Sewage Treatment Plant

    Science.gov (United States)

    Hephzibah, D.; Kumaran, P.; Saifuddin, N. M.

    2016-03-01

    This work elucidates the effects of pre-treatment of secondary thickened sludge (STS) for enhancement of biogas production that has great potential to generate energy for the utilization of the sewage treatment plant (STP) itself. Microwave pre-treatment has been adopted for this study. Experiment works have been designed and conducted to examine the effectiveness of continuous flow microwave pre-treatment on the solubility of STS, digestibility of STS and biogas production at a power level of 80 W for 5, 10 and 15 minutes. A few characteristics of the sewage sludge were monitored daily to identify the effect of pre-treatment on the sludge. The soluble chemical oxygen demand (SCOD)/total chemical oxygen demand (TCOD) ratio increased by 0.1, 1.0 and 1.8%, while the volatile fatty acids (VFA) concentration of the pre-treated sludge improved by 4.4, 5.1, 5.9% at the irradiation time of 5, 10 and 15 minutes, respectively at a microwave power level of 80 W. Besides that, the digestate also indicates that the pre-treated sludge undergoes efficient VS removal and TCOD removal after anaerobic digestion compared to the untreated sludge. Moreover, the biogas quantity increased by an average of 19.2, 24.1 and 32.2% in 5, 10 and 15 minutes irradiation time respectively compared to the untreated sludge. The additional quantity of biogas generated has shown a great potential for sustainable energy generation that can be utilized internally by the STP.

  18. A concise design o the irradiation of U-10Zr metallic fuel at a very low burnup

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Hai Bing; Zhou, Wei; Sun, Yong; Qian, Dazhi; Ma, Jimin; Leng, Jun; Huo, Hyoung; Wang, Shaohua [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang (China)

    2017-06-15

    In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

  19. Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel

    Science.gov (United States)

    Akie, H.; Sugo, Y.; Okawa, R.

    2003-06-01

    A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

  20. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

  2. Energy-saving and pollution-minimization continuous dyeing processes%印染厂连续染整节能减排工艺

    Institute of Scientific and Technical Information of China (English)

    马学亚; 冯森

    2011-01-01

    In order to save water and energy and reduce emissions, a senes of short processes are developed.Pretreatment processes include low alkali desizing and scouring and peroxide bleaching process, enzyme desizing and scouring and peroxide bleaching process, as well as peroxide desizing and scouring and peroxide bleaching process.Dyeing processes include saltfree continuous pad-steam process with reactive dyes, pad dyeing of cotton fabric with pigments, pad dyeing of polyester/cotton fabric with disperse dyes and pigmentS in one bath, as well as pad dyeing of polyester/cotton fabric in pale shade with disperse dyes.Water/oil/stain proofing and anti-static finishing in one bath is also introduced.%为了降低水耗和能耗,减少排污,开发了一系列短流程工艺,前处理包括低碱退煮氧漂、酶退煮氧漂和氧退煮氧漂工艺;染色包括活性染料无盐轧蒸工艺、纯棉涂料轧染、涤棉分散/涂料一浴轧染和浅色涤棉单分散轧染工艺;整理包括抗静电三防同浴工艺.

  3. Performance Evaluation of a Continuous Operation Adsorption Chiller Powered by Solar Energy Using Silica Gel and Water as the Working Pair

    Directory of Open Access Journals (Sweden)

    Hassan Zohair Hassan

    2014-10-01

    Full Text Available In the present study, dynamic analysis and performance evaluation of a solar-powered continuous operation adsorption chiller are introduced. The adsorption chiller uses silica gel and water as the working pair. The developed mathematical model represents the heat and mass transfer within the reactor coupled with the energy balance of the collector plate and the glass cover. Moreover, a non-equilibrium adsorption kinetic model is taken into account by using the linear driving force equation. The variation of solar radiation, wind speed, and atmospheric temperature along a complete cycle are considered for a more realistic simulation. Based on the case studied  and the baseline parameters, the chiller is found to acquire a coefficient of performance of 0.402. The average thermal efficiency of the solar collector is estimated to be 62.96% and the average total efficiency  approaches a value of 50.91%. Other performance parameters obtained are 363.8 W and 1.82 W/kg for the cooling capacity and the specific cooling power of the chiller, respectively. Furthermore, every 1 kg of silica gel inside the adsorption reactor produces a daily chilled water mass of 3 kg at a temperature of 10 ◦C. In addition, the cooling system harnesses 25.35% of the total available solar radiation and converts it to a cooling effect.

  4. The first three coefficients in the high temperature series expansion of free energy for simple potential models with hard-sphere cores and continuous tails.

    Science.gov (United States)

    Zhou, Shiqi; Solana, J R

    2013-08-08

    The first three coefficients of the high temperature series expansion (HTSE) of the Helmholtz free energy for a number of simple potential models with hard-sphere cores plus continuous tails are obtained for the first time from Monte Carlo simulations. The potential models considered include Square-well, Sutherland, attractive Yukawa, and triangle-well with different potential ranges, as well as a model potential qualitatively resembling the depletion potential in colloidal dispersions. The simulation data are used to evaluate performance of a recent coupling parameter series expansion (CPSE) in calculating for these coefficients, and a traditional macroscopic compressibility approximation (MCA) for the second-order coefficient only. A comprehensive comparison based on these coefficients from the two theoretical approaches and simulations enables one to conclude that (i) unlike one common experience that the widely used MCA usually underestimates the second-order coefficient, the MCA can both overestimate and underestimate the second-order coefficient, and worsens as the range of the potential decreases; and (ii) in contrast, the CPSE not only reproduce the trends in the density dependence of the perturbation coefficients, even the third one, observed in the simulations, but also the agreement is quantitative in most cases, and this clearly highlights the potential of the CPSE in providing accurate estimations for the higher-order coefficients, thus giving rise to an accurate higher-order HTSE.

  5. Energy

    Science.gov (United States)

    2003-01-01

    Canada, Britain, and Spain. We found that the energy industry is not in crisis ; however, U.S. government policies, laws, dollars, and even public...CEIMAT (Centro de Investagaciones Energeticas , Medioambeintales y Tecnologicas) Research and development Page 3 of 28ENERGY 8/10/04http://www.ndu.edu...meet an emerging national crisis (war), emergency (natural disaster), or major impact event (Y2K). Certain resources are generally critical to the

  6. An attempt to reproduce high burn-up structure by ion irradiation of SIMFUEL

    Energy Technology Data Exchange (ETDEWEB)

    Baranov, V.G. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation); Lunev, A.V., E-mail: AVLunev@mephi.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation); Reutov, V.F. [Joint Institute for Nuclear Research (JINR), Flerov Laboratory of Nuclear Reactions (FLNR), 141980 Dubna, Moscow Region (Russian Federation); Tenishev, A.V.; Isaenkova, M.G.; Khlunov, A.V. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation)

    2014-09-15

    Experiments in IC-100 and U-400 cyclotrons were conducted with SIMFUEL pellets (11.47 wt.% of fission products simulators) to reproduce some aspects of the long-term irradiation conditions in epithermal reactors. Pellets were irradiated with Xe{sup 16+}, Xe{sup 24+} and He{sup +} at energies ranging from 20 keV (He{sup +}) to 320 keV (Xe{sup 16+}) and 1–90 MeV (Xe{sup 24+}). Some samples were subsequently annealed to obtain larger grain sizes and to study defects recovery. The major microstructural changes consisted in grain sub-division observed on SEM and AFM images and change in composition registered by EPMA (pellets irradiated with 1–90 MeV Xe{sup 24+} ions at fluence of 5 × 10{sup 15} cm{sup −2}). Lattice distortion and increase in dislocation density is also noted according to X-ray data. At low energies and high fluences formation of bubbles (20 keV He{sup +} at 5.5 × 10{sup 17} cm{sup −2}) was observed. Grain sub-division exhibits full coverage of the grain body and preservation of former grain boundaries. The size of sub-grains depends on local dislocation density and changes from 200 nm to 400 nm along the irradiated surface. Beneath it the size ranges from 150 to 600 nm. Sub-grains are not observed in samples irradiated by low-energy ions even at high dislocation densities.

  7. Separation of metallic residues from the dissolution of a high-burnup BWR fuel using nitrogen trifluoride

    Energy Technology Data Exchange (ETDEWEB)

    McNamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mausolf, Edward J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Scheele, Randall D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-03-23

    Nitrogen trifluoride (NF3) was used to fluorinate the metallic residue from the dissolution of a high burnup, boiling water reactor fuel (~70 MWd/kgU). The metallic residue included the noble metal phase (containing ruthenium, rhodium, palladium, technetium, and molybdenum), and smaller amounts of zirconium, selenium, tellurium, and silver. Exposing the noble metal phase to 10% NF3 in argon between 400 and 550°C, removed molybdenum and technetium near 400°C as their volatile fluorides, and ruthenium near 500C as its volatile fluoride. The events were thermally and temporally distinct and the conditions specified are a recipe to separate these transition metals from each other and from the noble metal phase nonvolatile residue. Depletion of the volatile fluorides resulted in substantial exothermicity. Thermal excursion behavior was recorded under non-adiabatic, isothermal conditions that typically minimize heat release. Physical characterization of the metallic noble phase and its thermal behavior are consistent with high kinetic velocity reactions encouraged by the nanoparticulate phase or perhaps catalytic influences of the mixed platinum metals with nearly pure phase structure. Post-fluorination, only two phases were present in the residual nonvolatile fraction. These were identified as a nano-crystalline, metallic palladium cubic phase and a hexagonal rhodium trifluoride (RhF3) phase. The two phases were distinct as the sub-µm crystallites of metallic palladium were in contrast to the RhF3 phase, which grew from the parent nano-crystalline noble-metal phase during fluorination, to acicular crystals exceeding 20-µm in length.

  8. Development of a Membrane-Based Separation Process for the Continuous Enzymatic Saccharification of Lignocellulosic Biomass; NREL (National Renewable Energy Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Adhikari, B.; Pellegrino, J.; Stickel, J.; Sievers, J.

    2014-04-29

    We are currently evaluating the feasibility of performing continuous enzymatic hydrolysis of lignocellulosic biomass to product sugars using a membrane-assisted reaction/separation process. The overarching technical goals are to continuously remove the sugars—this lowers product feedback inhibition—retain and recycle active enzyme, and continuously recover the co-product of lignin. Experimental d d d currently evaluating the feasibility of performing continuous enzymatic hydrolysis of lignocellulosic biomass to product sugars using a membrane-assisted reaction/separation process. The overarching technical goals are to continuously remove the sugars -- this lowers product feedback inhibition --retain and recycle active enzyme, and continuously recover the co-product of lignin.

  9. Development of a Membrane-Based Separation Process for the Continuous Enzymatic Saccharification of Lignocellulosic Biomass; NREL (National Renewable Energy Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Adhikari, B.; Pellegrino, J.; Stickel, J.; Sievers, J.

    2014-04-29

    We are currently evaluating the feasibility of performing continuous enzymatic hydrolysis of lignocellulosic biomass to product sugars using a membrane-assisted reaction/separation process. The overarching technical goals are to continuously remove the sugars—this lowers product feedback inhibition—retain and recycle active enzyme, and continuously recover the co-product of lignin. Experimental d d d currently evaluating the feasibility of performing continuous enzymatic hydrolysis of lignocellulosic biomass to product sugars using a membrane-assisted reaction/separation process. The overarching technical goals are to continuously remove the sugars -- this lowers product feedback inhibition --retain and recycle active enzyme, and continuously recover the co-product of lignin.

  10. Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

    Science.gov (United States)

    Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

    2014-12-01

    Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

  11. Moderator poison design and burn-up calculations at the SNS

    Science.gov (United States)

    Lu, W.; Ferguson, P. D.; Iverson, E. B.; Gallmeier, F. X.; Popova, I.

    2008-06-01

    The spallation neutron source (SNS) at Oak Ridge National Laboratory was commissioned in April 2006. At the nominal operating power (1.4 MW), it will have thermal neutron fluxes approximately an order of magnitude greater than any existing pulsed spallation source. It thus brings a serious challenge to the lifetime of the moderator poison sheets. The SNS moderators are integrated with the inner reflector plug (IRP) at a cost of ˜$2 million a piece. A replacement of the inner reflector plug presents a significant drawback to the facility due to the activation and the operation cost. Although there are a lot of factors limiting the lifetime of the inner reflector plug, like radiation damage to the structural material and helium production of beryllium, the bottle-neck is the lifetime of the moderator poison sheets. Increasing the thickness of the poison sheet extends the lifetime but would sacrifice the neutronic performance of the moderators. A compromise is accepted at the current SNS target system which uses thick Gd poison sheets at a projected lifetime of 6 MW-years of operation. The calculations in this paper reveal that Cd may be a better poison material from the perspective of lifetime and neutronic performance. In replacing Gd, the inner reflector plug could reach a lifetime of 8 MW-years with ˜5% higher peak neutron fluxes at almost no loss of energy resolution.

  12. Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis

    CERN Document Server

    Onegin, M S

    2014-01-01

    New severe constraints on the variation of the fine structure constant have been obtained from reactor Oklo analysis in our previous work. We investigate here how these constraints confine the parameter of BSBM model of varying $\\alpha$. Integrating the coupled system of equations from the Big Bang up to the present time and taking into account the Oklo limits we have obtained the following margin on the combination of the parameters of BSBM model: $$ |\\zeta_m (\\frac{l}{l_{pl}})^2|<6\\cdot 10^{-7}, $$ where $l_{pl}=(\\frac{G\\hbar}{c^3})^{\\frac{1}{2}} \\approx 1.6 \\cdot 10^{-33}$ cm is a Plank length and $l$ is the characteristic length of the BSBM model. The natural value of the parameter $\\zeta_m$ - the fraction of electromagnetic energy in matter - is about $10^{-4}$. As a result it is followed from our analysis that the characteristic length $l$ of BSBM theory should be considerably smaller than the Plank length to fulfill the Oklo constraints on $\\alpha$ variation.

  13. Continuity theory

    CERN Document Server

    Nel, Louis

    2016-01-01

    This book presents a detailed, self-contained theory of continuous mappings. It is mainly addressed to students who have already studied these mappings in the setting of metric spaces, as well as multidimensional differential calculus. The needed background facts about sets, metric spaces and linear algebra are developed in detail, so as to provide a seamless transition between students' previous studies and new material. In view of its many novel features, this book will be of interest also to mature readers who have studied continuous mappings from the subject's classical texts and wish to become acquainted with a new approach. The theory of continuous mappings serves as infrastructure for more specialized mathematical theories like differential equations, integral equations, operator theory, dynamical systems, global analysis, topological groups, topological rings and many more. In light of the centrality of the topic, a book of this kind fits a variety of applications, especially those that contribute to ...

  14. Space heating savings via continuous building renovation up to 2050. Network for energy renovation; Varmebesparelse ved loebende bygningsrenovering frem til 2050. Netvaerk for energirenovering

    Energy Technology Data Exchange (ETDEWEB)

    Wittchen, K.B.; Kragh, J.

    2013-03-15

    The report presents analysis of net heat savings related to ongoing building renovations until 2050 if building parts are insulated according to requirements of the Building Regulations 2010 at the time when the building parts anyhow must undergo regular renovation and maintenance or replacement. The analysis is compared with an assessment of the effect of a 100% implementation of the energy requirements of BR10 for renovation / replacement of building components. This comparison provides the ultimate energy saving using more insulation in connection with the planned renovation and replacement. The underlying calculation model is built using data from the Danish Building and Housing Register and statistical data from the Energy Labeling Scheme for building insulation standard and heated area of roof, exterior wall, floor / slab and windows / doors. The report was prepared for the Danish Energy Agency and is targeted at participants in the network for energy renovation, especially the construction industry and agencies, and policy makers. (LN)

  15. Treatment of energy loss and multiple scattering in the context of track parameter and covariance matrix propagation in continuous material in the ATLAS experiment

    CERN Document Server

    Lund, E; Hughes, E W; Lopez Mateos, D; Salzburger, A; Strandlie, A

    2008-01-01

    In this paper we study the energy loss, its fluctuations, and the multiple scattering of particles passing through matter, with an emphasis on muons. In addition to the well-known Bethe-Bloch and Bethe-Heitler equations describing the mean energy loss from ionization and bremsstrahlung respectively, new parameterizations of the mean energy loss of muons from the direct e+e- pair production and photonuclear interactions are presented along with new estimates of the most probable energy loss and its fluctuations in the ATLAS calorimeters. Moreover, a new adaptive Highland/Moliere approach to finding the multiple scattering angle is taken to accomodate a wide range of scatterer thicknesses. Furthermore, tests of the muon energy loss, its fluctuations, and multiple scattering are done in the ATLAS calorimeters. The material effects described in this paper are all part of the simultaneous track and error propagation (STEP) algorithm of the common ATLAS tracking software.

  16. Paul Scherrer Institut Scientific Report 2002. Volume IV: Nuclear Energy and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Smith, B.; Gschwend, B. (eds.)

    2003-03-01

    Highlights in research and operation Projects established in previous years have yielded relevant and first-of-a-kind results, which have gained broad attention, both nationally and internationally, and which are presented in detail in this report. A few outstanding examples are cited below:Successful first measurements with highly active samples in LWR-PROTEUS Phase II (high burn-up fuel) have shown significant dependency of reactivity on burn-up, and increasing discrepancies between calculated and measured reactivity values with burn-up. As a consequence of these findings, the Swiss utilities wish to extend this phase. On-call calculations in the framework of the STARS project have been used to modify the feedwater system of the Leibstadt NPP. The modification has been subsequently confirmed during a turbine trip. An international consortium has been established for the ARTIST project (aerosol behaviour in the case of steam generator tube rupture). First tests showed higher aerosol retention than expected. The MEGAPIE project remains ongoing, and plans for post-irradiation examination (PIE) have now been established. Investigation of the leak which occurred in the LISOR loop in the Hot Lab has positively identified the cause of failure. PSI's contribution to the China Energy Technology Programme has been completed and documented. Among other results, the programme provided evidence for lower total costs (including externalities) by using 'clean coal' technologies. In parallel, and to assure continuation of the successful collaboration with the European research programmes, NES has participated in 27 'Expressions of Interest' for Integrated Projects and Networks of Excellence within the 6th EU Framework Programme. On the operational level, the year 2002 was marked by a series of significant events: The Federal Institutes of Technology, to which PSI belongs, have drawn up strategic plans for the years 2004-2007. The proposed PSI

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

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

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

  18. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    Energy Technology Data Exchange (ETDEWEB)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V. [Inst. of Radiobiology, Minsk Univ. (Belarus); Boulyga, S.F. [Inst. of Inorganic Chemistry and Analytical Chemistry, Johannes Gutenberg-Univ. Mainz, Mainz (Germany); Becker, J.S. [Central Div. of Analytical Chemistry, Research Centre Juelich, Juelich (Germany)

    2005-07-01

    An analytical method is described for the estimation of uranium concentrations, of {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10{sup -9}g/g to 2.0 x 10{sup -6}g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and the average value amounted to 9.4{+-}0.3 MWd/(kg U). (orig.)

  19. Lattice parameter changes associated with the rim-structure formation in high burn-up UO 2 fuels by micro X-ray diffraction

    Science.gov (United States)

    Spino, J.; Papaioannou, D.

    2000-10-01

    Radial variations of the lattice parameter and peak width of two high burn-up UO 2-fuels (67 and 80 GWd/tM) were measured by a specially developed micro-X-ray diffraction technique, allowing spectra acquisition with 30 μm spatial resolution. The results showed a significant but constant peak broadening, and a lattice parameter that increased towards the pellet edge and decreased again within the rim-zone. This lattice contraction coincided with other property changes in the rim region, i.e., porosity increase, hardness decrease and Xe depletion. In terms of local burn-ups, the lattice contraction followed the rate of the matrix Xe depletion measured by EMPA, exceeding greatly the contraction rate due to dissolved fission products. The observed behaviour can be equally explained by a saturation of single interstitials with subsequent recombination with excess vacancies, as by the saturation and enlargement of dislocation loops. The concentration and sizes of defects involved and their possible relation to the rim structure formation are discussed.

  20. Neutronic Study of Burnup, Radiotoxicity, Decay Heat and Basic Safety Parameters of Mono-Recycling of Americium in French Pressurised Water Reactors

    Directory of Open Access Journals (Sweden)

    Robert Bright Mawuko Sogbadji

    2017-03-01

    Full Text Available The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu. This work looks at reducing the actinides inventory nuclear fuel waste meant for permanent disposal. The uranium oxide fuel (UOX assembly, as in the open cycle system, was designed to reach a burnup of 46GWd/T and 68GWd/T using the MURE code. The MURE code is based on the coupling of a static Monte Carlo code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to address two different questions concerning the mono-recycling of americium (Am in present French pressurised water reactors (PWR. These are reduction of americium in the clear fuel cycle and the safe quantity of americium that can be introduced into mixed oxide (MOX as fuel. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of plutonium and addition of depleted uranium to reach burnups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. After 30 years of cooling in the repository, the spent UOX fuel required a higher concentration of Pu to be reprocessed into MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has a high radiotoxicity level, high mid-term residual heat and is a precursor for other long-lived isotopes. An innovative strategy would be to reprocess not only the plutonium from the UOX spent fuel but also the americium isotopes, which presently dominate the radiotoxicity of waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all americium. The main objective is to propose a ‘waiting strategy’ for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOX and

  1. Energy Technology Division research summary 1997.

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-21

    The Energy Technology Division provides materials and engineering technology support to a wide range of programs important to the US Department of Energy. As shown on the preceding page, the Division is organized into ten sections, five with concentrations in the materials area and five in engineering technology. Materials expertise includes fabrication, mechanical properties, corrosion, friction and lubrication, and irradiation effects. Our major engineering strengths are in heat and mass flow, sensors and instrumentation, nondestructive testing, transportation, and electromechanics and superconductivity applications. The Division Safety Coordinator, Environmental Compliance Officers, Quality Assurance Representative, Financial Administrator, and Communication Coordinator report directly to the Division Director. The Division Director is personally responsible for cultural diversity and is a member of the Laboratory-wide Cultural Diversity Advisory Committee. The Division's capabilities are generally applied to issues associated with energy production, transportation, utilization or conservation, or with environmental issues linked to energy. As shown in the organization chart on the next page, the Division reports administratively to the Associate Laboratory Director (ALD) for Energy and Environmental Science and Technology (EEST) through the General Manager for Environmental and Industrial Technologies. While most of our programs are under the purview of the EEST ALD, we also have had programs funded under every one of the ALDs. Some of our research in superconductivity is funded through the Physical Research Program ALD. We also continue to work on a number of nuclear-energy-related programs under the ALD for Engineering Research. Detailed descriptions of our programs on a section-by-section basis are provided in the remainder of this book. This Overview highlights some major trends. Research related to the operational safety of commercial light water

  2. Continuation calculus

    Directory of Open Access Journals (Sweden)

    Bram Geron

    2013-09-01

    Full Text Available Programs with control are usually modeled using lambda calculus extended with control operators. Instead of modifying lambda calculus, we consider a different model of computation. We introduce continuation calculus, or CC, a deterministic model of computation that is evaluated using only head reduction, and argue that it is suitable for modeling programs with control. It is demonstrated how to define programs, specify them, and prove them correct. This is shown in detail by presenting in CC a list multiplication program that prematurely returns when it encounters a zero. The correctness proof includes termination of the program. In continuation calculus we can model both call-by-name and call-by-value. In addition, call-by-name functions can be applied to call-by-value results, and conversely.

  3. 阶梯型脉冲电压诱导连续能量质子谱数值仿真%Numerical simulation of proton generation with continuous energy spectrum by pulse voltage with discrete steps

    Institute of Scientific and Technical Information of China (English)

    石经纬; 汪志健; 巩春志; 田修波; 杨士勤

    2011-01-01

    The performance degradation of spacecraft thermal control coatings irradiated by protons is generally investigated by using protons with the same energy in ground testing while the energy of protons is in succession in space. The irradiation e-quivalence of the two kinds of protons is still not well understood. In this paper, a method of producing protons with continuous energy by plasma sheath acceleration using pulse voltage with discrete steps is proposed for better analysis of irradiation equivalence. The dose-energy distribution on the sample is numerically investigated by particle-in-cell(PIC) method. The characteristics of dose-energy distribution and the formation mechanism of protons with continuous energy are then discussed. The results show that protons with continuous energy can be realized utilizing pulse voltage with discrete steps, as the energy of protons irradiating the sample overlaps between two adjacent 1 μs periods, and the energy of protons produced is closely related to the voltage on the sample in every 1 μs period. Moreover, the number of protons irradiating the sample may decrease if the proton energy increases.%热控涂层质子辐照的地面模拟研究中采用单一能量质子替代空间能量连续分布的质子,连续能量质子谱是其等效性研究的关键.提出了采用阶梯型脉冲负偏压鞘层加速技术在一个脉冲宽度内获得连续能量质子谱的方法,并利用质点网格法对所获得质子谱的剂量-能量关系进行了数值仿真研究,分析了连续能量质子谱的剂量-能量分布特征及连续能量质子谱的形成过程.结果表明:阶梯型脉冲负偏压鞘层加速能够产生连续能量的质子谱,连续谱是每微秒区间入射到样品的质子叠加而成的,且每个区间所产生质子的能量与该区间电压值相对应,连续谱中,随着质子能量的增加,其剂量总体上呈现下降的趋势.

  4. A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy; Um modelo estocastico de simulacao neutronica considerando o espectro e propriedades nucleares com dependencia continua de energia

    Energy Technology Data Exchange (ETDEWEB)

    Camargo, Dayana Queiroz de

    2011-01-15

    This thesis has developed a stochastic model to simulate the neutrons transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using Monte Carlo method for the propagation of neutrons in different environment. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational processing time introduced the variable control volume along the (pseudo-) periodic boundary conditions in order to overcome this problem. The choice of class physical Monte Carlo is due to the fact that it can decompose into simpler constituents the problem of solve a transport equation. The components may be treated separately, these are the propagation and interaction while respecting the laws of energy conservation and momentum, and the relationships that determine the probability of their interaction. We are aware of the fact that the problem approached in this thesis is far from being comparable to building a nuclear reactor, but this discussion the main target was to develop the Monte Carlo model, implement the code in a computer language that allows extensions of modular way. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude the need to consider the energy dependence, i.e. an spectral effective multiplication factor should be introduced each energy group separately. (author)

  5. China Needs to Continue to Develop Nuclear Energy in the New Period%新时期我国继续发展核电的必要性

    Institute of Scientific and Technical Information of China (English)

    陈方强; 王青松

    2012-01-01

    With an analysis on the present situation of energy structure and nuclear power development in China and demonstration of the advantages of nuclear power, this study discusses the actual need for nuclear power in terms of energy saving and emission reduction, energy structure optimization, satisfaction of the growing energy demand and stable power supply, industrial development, national security and improvement of the comprehensive national power. It concludes that development of high-efficiency nuclear power in China is of great significance and an inevitable choice of China.%分析论证了我国能源结构现状、核电发展现状及核电优点;从节能减排和能源结构优化、满足能源需求增长和电力稳定供应、产业经济发展、国家安全需要和综合国力具体体现等方面分析了我国安全高效发展核电的现实需要,认为高效发展核电仍是我国的必然选择。

  6. Burn-up Function of Fuel Management Code for Aqueous Homogeneous Reactors and Its Validation%溶液堆物理计算程序FMCAHR燃耗功能及其验证

    Institute of Scientific and Technical Information of China (English)

    汪量子; 姚栋; 王侃

    2011-01-01

    介绍了FMCAHR程序的燃耗计算模型及流程,并使用燃耗基准题和DRAGON程序对燃耗计算结果进行验证.验证结果表明,FMCAHR燃耗计算功能的准确性较高,适用于溶液堆的燃耗计算分析.%Fuel Management Code for Aqueous Homogeneous Reactors(FMCAHR)is developed based on the Monte Carlo transport method,to analyze the physics characteristics of aqueous homogeneous reactors. FMCAHR has the ability of doing resonance treatment,searching for critical rod heights,thermal hydraulic parameters calculation,radiolytic-gas bubbles' calculation and burn-up calculation. This paper introduces the theory model and scheme of its bum-up function,and then compares its calculation results with benchmarks and with DRAGON'S burn-up results,which confirms its burn-up computing precision and its applicability in the burn-up calculation and analysis for aqueous solution reactors.

  7. Development of Burnup Calculation Function in Reactor Monte Carlo Code RMC%堆用蒙卡程序燃耗计算功能开发

    Institute of Scientific and Technical Information of China (English)

    佘顶; 王侃; 余纲林

    2012-01-01

    This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua university of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.%堆用蒙卡程序(RMC)是由清华大学工程物理系REAL实验室自主开发的用于反应堆物理分析的中子输运蒙卡程序,本文主要介绍其燃耗计算功能的开发与验证.RMC的燃耗计算功能具有的特点:内部耦合ORIGEN,相比于外耦合方式,更加灵活和高效;使用基于能谱的单群截面统计方法,可在保证精度的前提下,显著提高计算效率;采取预估修正和中点近似等多种燃耗步策略,减小大燃耗步长时的计算误差.通过计算压水堆栅元、沸水堆组件、快堆等一系列基准题和算例,验证了RMC燃耗计算的正确性和速度优势.

  8. Continuity and internal properties of Gulf Coast sandstones and their implications for geopressured energy development. Annual report, November 1, 1980-October 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Morton, R.A.; Ewing, T.E.; Tyler, N.

    1982-06-01

    Systematic investigation, classification, and differentiation of the intrinsic properties of genetic sandstone units that typify many geopressured geothermal aquifers and hydrocarbon reservoirs of the Gulf Coast region are provided. The following are included: structural and stratigraphic limits of sandstone reservoirs; characteristics and dimensions of Gulf Coast Sandstones; fault compartment areas; comparison of production and geologic estimates of aquifer volume; geologic setting and reservoir characteristics, wells of opportunity; internal properties of sandstones and implications for geopressured energy development. (MHR)

  9. The MOX fuel behaviour test IFA-597.4/.5/.6; Thermal and gas release data to a burn-up of 25 MWd/kgMOX

    Energy Technology Data Exchange (ETDEWEB)

    Tolonen, Pekka; Pihlatie, Mikko; Fujii, Hajime

    2001-02-15

    Responding to the needs of the member organisations, a research programme of MOX fuel has been established in the joint programme of the Halden Reactor Project. IFA-597.4, IFA- 597.5, and IFA-597.6, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, have been irradiated in the Halden Reactor since July 1997. The objective of the test series is study the thermal and fission gas release behaviour of MOX fuel, and to explore differences in performance between solid and hollow pellets. One of the rods has mainly solid pellets, while the other contains only hollow pellets. Both rods have an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter is 9.5 mm, and the initial fuel-cladding gap 180 mum. The rods are being periodically uprated to study the fission gas release onset of MOX fuel. The first uprating was performed at approx 10 MWd/kgMOX resulting in significant gas release in both rods. In order to accumulate fission gas in the fuel matrix instead of releasing it, four UO{sub 2} rods were added to the rig at approx 13.5 MWd/kgMOX to suppress the linear heat rate of the MOX rods. Consequently, no gas release occurred during this low power operation. The UO{sub 2} rods were removed at approx 20.5 MWd/kgMOX, resulting in a power increase and significant gas release in both rods. The burnup of the rods has reached approx 25 MWd/kgMOX as of the end of October 2000. The following loading will operate at low power until late 2001 to avoid fission gas release. The target burnup of the test is 60 MWd/kgMOX. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-09

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

  11. Continuous fat oxidation in acetyl–CoA carboxylase 2 knockout mice increases total energy expenditure, reduces fat mass, and improves insulin sensitivity

    Science.gov (United States)

    Choi, Cheol Soo; Savage, David B.; Abu-Elheiga, Lutfi; Liu, Zhen-Xiang; Kim, Sheene; Kulkarni, Ameya; Distefano, Alberto; Hwang, Yu-Jin; Reznick, Richard M.; Codella, Roberto; Zhang, Dongyan; Cline, Gary W.; Wakil, Salih J.; Shulman, Gerald I.

    2007-01-01

    Acetyl–CoA carboxylase 2 (ACC)2 is a key regulator of mitochondrial fat oxidation. To examine the impact of ACC2 deletion on whole-body energy metabolism, we measured changes in substrate oxidation and total energy expenditure in Acc2−/− and WT control mice fed either regular or high-fat diets. To determine insulin action in vivo, we also measured whole-body insulin-stimulated liver and muscle glucose metabolism during a hyperinsulinemic–euglycemic clamp in Acc2−/− and WT control mice fed a high-fat diet. Contrary to previous studies that have suggested that increased fat oxidation might result in lower glucose oxidation, both fat and carbohydrate oxidation were simultaneously increased in Acc2−/− mice. This increase in both fat and carbohydrate oxidation resulted in an increase in total energy expenditure, reductions in fat and lean body mass and prevention from diet-induced obesity. Furthermore, Acc2−/− mice were protected from fat-induced peripheral and hepatic insulin resistance. These improvements in insulin-stimulated glucose metabolism were associated with reduced diacylglycerol content in muscle and liver, decreased PKCθ activity in muscle and PKCε activity in liver, and increased insulin-stimulated Akt2 activity in these tissues. Taken together with previous work demonstrating that Acc2−/− mice have a normal lifespan, these data suggest that Acc2 inhibition is a viable therapeutic option for the treatment of obesity and type 2 diabetes. PMID:17923673

  12. Energy efficiency project: continuous rolling mill profiles plant of Barra Mansa Votorantim Siderurgia; Projeto de eficiencia energetica: laminador continuo de perfis da Usina Barra Mansa Votorantim Siderurgia

    Energy Technology Data Exchange (ETDEWEB)

    Decnop, Luiz Eduardo Machado [LD Engenharia Ltda, Belo Horizonte, MG (Brazil); Franca, Rafael Latini; Silva, Bruno Jose Carvalho; Antunes, Raphael Jose Simoes [Votorantim Siderurgia, Volta Redonda, RJ (Brazil)

    2011-12-21

    This work shows the development and implementation of a plan to increase energy efficiency and consequent reduction in production costs of profiles. The plant has a varied mix, large differences among products (1 kg/m to 30 kg/m) that resulting in a large gaps between bars and high level of interruption. It was necessary to reduce the impacts of fixed charges that weighed heavily in the cost. The methodology was to identify, among the main elements contributing to the consumption of electricity, what is the real need of each input as the quality, quantity, intensity and permanence, to perform strictly their intended function in the process. Then the comparison of these results with the engineering solutions adopted in the current design of the equipment. Thus was taken to waste all resources available beyond the strict needs. This ranking conducted the study to optimize the delivery of each resource taking into account the difficulty and risk of possible technical solutions, taking as a premise to develop definitive solutions with full automation of security functions to ensure stability and security of gains. The results were the reduction of 147 kWh/t to 105 kWh/t in the LCP (-37%), and a 50% reduction in energy expended in the WTP. (author)

  13. Analytical continuous slowing down model for nuclear reaction cross-section measurements by exploitation of stopping for projectile energy scanning and results for 13C(3He,α)12C and 13C(3He,p)15N

    Science.gov (United States)

    Möller, S.

    2017-03-01

    Ion beam analysis is a set of precise, calibration free and non-destructive methods for determining surface-near concentrations of potentially all elements and isotopes in a single measurement. For determination of concentrations the reaction cross-section of the projectile with the targets has to be known, in general at the primary beam energy and all energies below. To reduce the experimental effort of cross-section measurements a new method is presented here. The method is based on the projectile energy reduction when passing matter of thick targets. The continuous slowing down approximation is used to determine cross-sections from a thick target at projectile energies below the primary energy by backward calculation of the measured product spectra. Results for 12C(3He,p)14N below 4.5 MeV are in rough agreement with literature data and reproduce the measured spectra. New data for reactions of 3He with 13C are acquired using the new technique. The applied approximations and further applications are discussed.

  14. Environmental assessment for the Satellite Power System (SPS): studies of honey bees exposed to 2. 45 GHz continuous-wave electromagnetic energy

    Energy Technology Data Exchange (ETDEWEB)

    Gary, N E; Westerdahl, B B

    1980-12-01

    A system for small animal exposure was developed for treating honey bees, Apis mellifera L., in brood and adult stages, with 2.45 GHz continuous wave microwaves at selected power densities and exposure times. Post-treatment brood development was normal and teratological effects were not detected at exposures of 3 to 50 mw/cm/sup 2/ for 30 minutes. Post-treatment survival, longevity, orientation, navigation, and memory of adult bees were also normal after exposures of 3 to 50 mw/cm/sup 2/ for 30 minutes. Post-treatment longevity of confined bees in the laboratory was normal after exposures of 3 to 50 mw/cm/sup 2/ for 24 hours. Thermoregulation of brood nest, foraging activity, brood rearing, and social interaction were not affected by chronic exposure to 1 mw/cm/sup 2/ during 28 days. In dynamic behavioral bioassays the frequency of entry and duration of activity of unrestrained, foraging adult bees was identical in microwave-exposed (5 to 40 mw/cm/sup 2/) areas versus control areas.

  15. Alternative energies; Energies alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Bonal, J.; Rossetti, P

    2007-07-01

    The earth took millions years to made the petroleum, the gas the coal and the uranium. Only a few centuries will be needed to exhaust these fossil fuels and some years to reach expensive prices. Will the wold continue on this way of energy compulsive consumption? The renewable energies and some citizen attitudes are sufficient to break this spiral. This book proposes to discuss these alternative energies. It shows that this attitude must be supported by the government. It takes stock on the more recent information concerning the renewable energies. it develops three main points: the electricity storage, the housing and the transports. (A.L.B.)

  16. Sizing design for continuous-operation off-grid wind power generation system with battery energy storage%离网自治型蓄电池储能风力发电系统规格设计

    Institute of Scientific and Technical Information of China (English)

    卢闻州; 惠晶

    2016-01-01

    Based on the traditional configuration of the off-grid wind generation system with battery energy storage, this paper provides an energy control method and a complete system sizing scheme, mainly including wind turbine generator and battery, to achieve continuous operation without an excess energy consuming device and decrease the system size and cost.In this paper, Hong Kong and Denmark are selected to do the case studies, in where system si-zes are designed by using the proposed scheme.Simulation results show that, based on continuous operation as the precondition, the designed sizes of wind turbine generator and battery are the required minimum ones, which can make both Hong Kong and Denmark achieve yearly energy balance between supply and demand.Therefore, simulation results show the validity and rationality of this sizing design scheme for continuous-operation off-grid wind power gener-ation system with battery energy storage.%在传统蓄电池储能离网风力发电系统结构基础上,去除泻荷装置,在保证风力发电系统自治、连续不间断运行的前提下,分析系统能量控制方法,给出系统中各组成元件规格的设计方法,主要包括风电机组尺寸及蓄电池容量的设计,以期减小自治离网系统尺寸,降低系统成本。选取香港和丹麦两地进行案例研究,依据所提规格设计方法进行系统尺寸设计,仿真结果表明:两地系统都实现了全年能量供需平衡,设计出的风电机组尺寸和蓄电池容量为两地实现连续、自治、不间断运行前提下的最小配置,从而验证了所提蓄电池储能离网自治型风力发电系统规格设计方法的合理性与有效性。

  17. Optimization of FRAM precision for isotopic measurements on large samples of low-burnup PuO2

    Energy Technology Data Exchange (ETDEWEB)

    Vo, Duc T [Los Alamos National Laboratory; Wenz, Tracy R [Los Alamos National Laboratory; Sampson, Thomas E [Los Alamos National Laboratory

    2009-01-01

    The gamma ray spectrum of plutonium contains measurable gamma rays ranging in energy from 60 keV to above 1 MeV. The FRAM gamma ray isotopic analysis code can analyze data from all types of HPGe detectors in this energy range typically using planar detectors in the energy range 60-210 keV or 120-451 keV and using coaxial detectors in the energy ranges 120-451 keV or 200-1001 keV. The statistical measurement precision depends upon the detector/energy range combination as well as the characteristics of the sample and any addition filters. In this paper we carry out the optimization of measurement precision for the important case of a multi-kg sample of low bumup PuO{sub 2} contained in a DOE 3013 Standard-compatible long-teml storage container.

  18. 78 FR 21245 - Continuity of Operations Plan

    Science.gov (United States)

    2013-04-10

    ... Federal Energy Regulatory Commission 18 CFR Part 376 Continuity of Operations Plan AGENCY: Federal Energy... Continuity of Operations Plan regulations to revise its hierarchy of delegation of Commission authority... Rule revises the Commission's Continuity of Operations Plan (COOP) regulations to incorporate...

  19. Energies; Energies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    In the framework of the National Debate on the energies in a context of a sustainable development some associations for the environment organized a debate on the nuclear interest facing the renewable energies. The first part presents the nuclear energy as a possible solution to fight against the greenhouse effect and the associated problem of the wastes management. The second part gives information on the solar energy and the possibilities of heat and electric power production. A presentation of the FEE (French wind power association) on the situation and the development of the wind power in France, is also provided. (A.L.B.)

  20. Protein design using continuous rotamers.

    Directory of Open Access Journals (Sweden)

    Pablo Gainza

    2012-01-01

    Full Text Available Optimizing amino acid conformation and identity is a central problem in computational protein design. Protein design algorithms must allow realistic protein flexibility to occur during this optimization, or they may fail to find the best sequence with the lowest energy. Most design algorithms implement side-chain flexibility by allowing the side chains to move between a small set of discrete, low-energy states, which we call rigid rotamers. In this work we show that allowing continuous side-chain flexibility (which we call continuous rotamers greatly improves protein flexibility modeling. We present a large-scale study that compares the sequences and best energy conformations in 69 protein-core redesigns using a rigid-rotamer model versus a continuous-rotamer model. We show that in nearly all of our redesigns the sequence found by the continuous-rotamer model is different and has a lower energy than the one found by the rigid-rotamer model. Moreover, the sequences found by the continuous-rotamer model are more similar to the native sequences. We then show that the seemingly easy solution of sampling more rigid rotamers within the continuous region is not a practical alternative to a continuous-rotamer model: at computationally feasible resolutions, using more rigid rotamers was never better than a continuous-rotamer model and almost always resulted in higher energies. Finally, we present a new protein design algorithm based on the dead-end elimination (DEE algorithm, which we call iMinDEE, that makes the use of continuous rotamers feasible in larger systems. iMinDEE guarantees finding the optimal answer while pruning the search space with close to the same efficiency of DEE.Software is available under the Lesser GNU Public License v3. Contact the authors for source code.

  1. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  2. Calculation of Effect of Burnup History on Spent Fuel Reactivity Based on CASMO5%基于CASMO5的燃耗历史对乏燃料反应性的影响计算

    Institute of Scientific and Technical Information of China (English)

    李晓波; 夏兆东; 朱庆福

    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 histo‐ry 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 0.002 3 to k∞ ,and taking the operating history of specified power without shutdown during cycle and between cycles plus a bounding margin of 0.004 5 to k∞ can meet the bounding principle of burnup credit .%基于乏燃料贮存领域常用的锕系加裂变产物(APU‐2)级燃耗信任制,应用二维组件燃耗计算程序CASMO5,计算了燃耗过程中功率密度和运行历史对乏燃料 k∞的影响。结果表明:燃耗计算中,选择堆芯额定功率对应的平均功率密度,同时k∞附加0.0023的包络裕度,运行历史选择循环内及循环间无停堆额定功率运行,同时 k∞附加0.0045的包络裕度,可满足燃耗信任制中包络性原则。

  3. Development of Burnup Calculation Code for Pebble-bed High Temperature Reactor at Equilibrium State%球床高温堆平衡态燃耗计算程序的开发

    Institute of Scientific and Technical Information of China (English)

    朱贵凤; 邹杨; 李明海; 严睿; 彭红花; 徐洪杰

    2015-01-01

    The burnup calculation code PBRE coupling MCNP5 and ORIGEN2 was developed for pebble‐bed high temperature reactor at equilibrium state ,and it can be used to analyze the neutronic performance of equilibrium core .The iteration method was optimized in order to save Monte Carlo calculation time ,and the convergence can be reached in 10 iterative steps .The average discharged burnup for HTR‐10 is consistent with literature ,and it indicates that the PBRE is suitable to analyze the burnup for pebble‐bed reactor at equilibrium state .%基于MCNP5和ORIGEN2耦合方法,开发了平衡态下球床高温堆的燃耗计算程序PBRE ,用于堆的性能价值分析。为节省蒙特卡罗计算时间,对迭代收敛的方法进行优化,使之可在10个迭代步内收敛。使用PBRE对清华大学H T R‐10进行建模计算,得到的平均卸料燃耗深度与文献报道值一致,表明PBRE程序适用于球床堆平衡态的燃耗分析。

  4. Chemical states of fission products in irradiated (U{sub 0.3}Pu{sub 0.7})C{sub 1+x} fuel at high burn-ups

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Renu [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: arenu@barc.gov.in; Venugopal, V. [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel (U{sub 0.3}Pu{sub 0.7})C{sub 1+x}, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  8. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    Science.gov (United States)

    Brémier, S.; Inagaki, K.; Capriotti, L.; Poeml, P.; Ogata, T.; Ohta, H.; Rondinella, V. V.

    2016-11-01

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15-20 μm and migration of cladding elements to the fuel.

  9. On the oxidation state of UO{sub 2} nuclear fuel at a burn-up of around 100MWd/kgHM

    Energy Technology Data Exchange (ETDEWEB)

    Walker, C.T. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany)]. E-mail: clive.walker@itu.fzk.de; Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Papaioannou, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Winckel, S. Van [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Goll, W. [Framatome ANP GmbH, P.O. Box 3223, D-91050 Erlangen (Germany); Manzel, R. [Framatome ANP GmbH, P.O. Box 3223, D-91050 Erlangen (Germany)

    2005-10-15

    Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102MWd/kgHM are presented. The local {delta}G-bar (O{sub 2}) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991+/-0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the {delta}G-bar (O{sub 2}) of the fuel increased from -370kJmol{sup -1} at r/r{sub 0}=0.1 to -293kJmol{sup -1} at r/r{sub 0}=0.975. Thus, the {delta}G-bar (O{sub 2}) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

  10. Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium.

    Science.gov (United States)

    Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

  11. On the oxidation state of UO 2 nuclear fuel at a burn-up of around 100 MWd/kgHM

    Science.gov (United States)

    Walker, C. T.; Rondinella, V. V.; Papaioannou, D.; Winckel, S. Van; Goll, W.; Manzel, R.

    2005-10-01

    Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102 MWd/kgHM are presented. The local Δ G bar (O2) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991 ± 0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the Δ G bar (O2) of the fuel increased from -370 kJ mol-1 at r/r0 = 0.1 to -293 kJ mol-1 at r/r0 = 0.975. Thus, the Δ G bar (O2) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

  12. Continuous Descent Operations using Energy Principles

    NARCIS (Netherlands)

    De Jong, P.M.A.

    2014-01-01

    During today’s aircraft descents, Air Traffic Control (ATC) commands aircraft to descend to specific altitudes and directions to maintain separation and spacing from other aircraft. When the aircraft is instructed to maintain an intermediate descent altitude, it requires engine thrust to maintain spe

  13. Effect of hydraulic bending roll on vibration energy of a hot continuous rolling mill%液压弯辊控制参数对热连轧机振动能量影响研究

    Institute of Scientific and Technical Information of China (English)

    闫晓强; 么爱东; 刘克飞

    2016-01-01

    In recent years,hot continuous rolling mill vibration problems are more prominent,many enterprises urgently need to solve this problem.Here,the vibration problem of a hot continuous rolling mill was monitored online. Then,a coupled dynamic model of a rolling mill vertical system and a hydraulic bending roll system was built.According to the actual mill parameters.MATLAB was used to do a simulation study.The effects of the control performance of the hydraulic bending rolls system on the rollers system vibration energy were analyzed by changing the controller parameters of the hydraulic bending roll system.The results provided one of effective measures for suppressing vibrations of hot continuous rolling mills.%近年来,热连轧机振动问题显得更加突出和复杂化,众多企业迫切需要解决这一难题。首先对某热连轧机振动现象进行在线监测;然后依据轧机实际参数建立液压弯辊系统和轧机垂直系统的耦合动力学模型,利用 MATLAB进行了仿真研究,通过改变液压弯辊系统中控制器参数,获得液压弯辊的控制性能对辊系振动能量的影响,实践表明这是有效抑制振动的措施之一。

  14. Monte Carlo simulations to calculate energy doses in a cow after continuous ingestion of CS 137 and K 40; Monte-Carlo-Simulationen zur Berechnung der Energiedosis in einem Rind nach kontinuierlicher Aufnahme von CS 137 und K 40

    Energy Technology Data Exchange (ETDEWEB)

    Pichl, E. [Technische Univ. Graz (Austria). Inst. fuer Medizintechnik; Rabitsch, H. [Technische Univ. Graz (Austria). Arbeitsgebiet Strahlenphysik

    2009-07-01

    Currently ICRP (International Commission on Radiological Protection) develops a new recommendation to estimate the natural radiation exposure of an agreed set of animals and reference plants. For estimating effective dose in humans and animals, the incorporated activities of natural and artificial radionuclides in body tissues and contents of the digestive system have to be known. It was the aim of this investigation to calculate energy doses caused by Cs 137 and K 40 in the reproductive organs (uterus, ovaries) of a cow. During its whole lifetime from 1986 to 1992, the cow incorporated continuously Cs 137 which was due to the fallout following the Chernobyl accident. K 40 occurs naturally in the cow's fodder. The cow was born in a highly contaminated region of Styria, Austria, and was infertile since 1990. The activities of Cs 137 and K 40 in the cow's fodder and in tissues, organs and contents of the digestive system of the carcass were measured simultaneously with the help of semiconductor detectors. To calculate the specific absorbed fractions by means of the Monte Carlo code MCNP, an appropriate simulation model for the reproductive organs and their surrounding tissues was developed. The contents of rectum and urinary bladder account for the main part of the energy dose in the reproductive organs. Comparison of our results with data from other investigations showed, that lifetime accumulation of Cs 137 and K 40 was too low to cause radiation inferred infertility. (orig.)

  15. 原料性质对连续重整装置能耗的影响%INFLUENCE OF FEED PROPERTIES ON THE ENERGY CONSUMPTION OF CONTINUOUS CATALYTIC REFORMING

    Institute of Scientific and Technical Information of China (English)

    王弘历; 龚燕; 王广河; 郭彦

    2012-01-01

    为了更加合理地反映原料性质对连续重整装置能耗的影响,以KBC公司的REF-SIM反应动力学模型和Petro-SIM流程模拟软件为工具,基于对重整反应过程的分析和数学回归方法,提出了一种新的关联原料性质与装置能耗的原料评价指标R(A/(N+P)).结果表明,与芳烃潜含量、芳构化指数等常用原料评价指标相比,原料表征指标R与装置能耗呈现较好的关联性,在相同产品辛烷值要求下,可以作为不同连续重整装置之间能效对标的基础评价参数.%With REF-SIM kinetics reaction model of continuous catalytic reforming (CCR) and Petro-SIM process simulation software, a new index, R(the ratio of aromaticsmass fraction to the mass fraction of naphthene and paraffin in the feed), of feed properties associated with unit energy consumption was proposed based on the analysis of reforming reaction process and mathematical regression method. It shows a good correlation between the new feed index R and unit energy consumption, better than that of the conventional feed index, potential aromatic content or aromatization index. Under the same octane requirement of product, the new feed index R can be used as a basic evaluation parameter when benchmarking energy consumption of different CCR units.

  16. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  17. Entanglement Continuous Unitary Transformations

    CERN Document Server

    Sahin, S; Orus, R

    2016-01-01

    Continuous unitary transformations are a powerful tool to extract valuable information out of quantum many-body Hamiltonians, in which the so-called flow equation transforms the Hamiltonian to a diagonal or block-diagonal form in second quantization. Yet, one of their main challenges is how to approximate the infinitely-many coupled differential equations that are produced throughout this flow. Here we show that tensor networks offer a natural and non-perturbative truncation scheme in terms of entanglement. The corresponding scheme is called "entanglement-CUT" or eCUT. It can be used to extract the low-energy physics of quantum many-body Hamiltonians, including quasiparticle energy gaps. We provide the general idea behind eCUT and explain its implementation for finite 1d systems using the formalism of matrix product operators, and we present proof-of-principle results for the spin-1/2 1d quantum Ising model in a transverse field. Entanglement-CUTs can also be generalized to higher dimensions and to the thermo...

  18. Entanglement continuous unitary transformations

    Science.gov (United States)

    Sahin, Serkan; Schmidt, Kai Phillip; Orús, Román

    2017-01-01

    Continuous unitary transformations are a powerful tool to extract valuable information out of quantum many-body Hamiltonians, in which the so-called flow equation transforms the Hamiltonian to a diagonal or block-diagonal form in second quantization. Yet, one of their main challenges is how to approximate the infinitely-many coupled differential equations that are produced throughout this flow. Here we show that tensor networks offer a natural and non-perturbative truncation scheme in terms of entanglement. The corresponding scheme is called “entanglement-CUT” or eCUT. It can be used to extract the low-energy physics of quantum many-body Hamiltonians, including quasiparticle energy gaps. We provide the general idea behind eCUT and explain its implementation for finite 1d systems using the formalism of matrix product operators. We also present proof-of-principle results for the spin-(1/2) 1d quantum Ising model and the 3-state quantum Potts model in a transverse field. Entanglement-CUTs can also be generalized to higher dimensions and to the thermodynamic limit.

  19. 多群蒙卡输运与点燃耗耦合程序系统TRITON基准验证%Benchmark Verification of Multi-group Monte Carlo Transport and Point-Burnup Codes Coupling System TRITON

    Institute of Scientific and Technical Information of China (English)

    武祥; 若夕子; 于涛; 谢金森; 陈昊威

    2014-01-01

    TRITON couples multi group Monte Carlo Transport code KENO V. a and point-burnup code ORIGEN-S. It features adaptability on complex geometries,flexible processing ability on cross section and rapid calculating speed. Based on the thorium-based fuel cell benchmark of Idaho National Laboratory ( INL) ,the verification on TRITON burnup calcu-lation was performed,which showed good coincidence with the result of MOCUP code by INL. Furthermore, the results of burnup isotopes selection schemes in TRITON showed that,for thorium based fuel,only important nuclides on Th-U cycle was included,correct results can be obtained by TRITON. Conclusions in the present paper will support further applications of TRITON.%TRITON程序系统耦合了多群蒙特卡罗输运程序KENO V. a与点燃耗程序ORIGEN-S,具有几何适应性强、截面处理能力灵活、计算速度快等显著特点.本文基于爱达荷国家实验室( INL)钍基燃料元件燃耗基准题,开展了TRITON程序燃耗功能的验证,结果与INL采用MOCUP程序给出的结果吻合很好.同时,燃耗核素选取对TRITON计算结果的影响分析表明对于钍基燃料,只有在考虑Th-U循环重要核素的前提下,TRITON才能给出正确结果.上述结论为TRITON程序的应用奠定了基础.

  20. Renewable Energy

    DEFF Research Database (Denmark)

    Sørensen, Bent Erik

    Bent Sorensen’s Renewable Energy: Physics, Engineering, Environmental Impacts, Economics and Planning, Fifth Edition, continues the tradition by providing a thorough and current overview of the entire renewable energy sphere. Since its first edition, this standard reference source helped put...... renewable energy on the map of scientific agendas. Several renewable energy solutions no longer form just a marginal addition to energy supply, but have become major players, with the promise to become the backbone of an energy system suitable for life in the sustainability lane. This volume is a problem...... structured around three parts in order to assist readers in focusing on the issues that impact them the most for a given project or question. PART I covers the basic scientific principles behind all major renewable energy resources, such as solar, wind, and biomass. PART II provides in-depth information...

  1. Energy: nuclear energy; Energies: l'energie nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Lung, M. [Societe Generale pour les Techniques Nouvelles (SGN), 78 - Saint-Quentin-en-Yvelines (France)

    2000-11-01

    Convinced that the nuclear energy will be the cleaner, safer, more economical and more respectful of the environment energy of the future, the author preconizes to study the way it can be implemented, to continue to improve its production, to understand its virtues and to better inform the public. He develops this opinion in the presentation of the principal characteristics of the nuclear energy: technology, radioactive wastes, radiation protection, the plutonium, the nuclear accidents, the proliferation risks, the economics and nuclear energy and competitiveness, development and sustainability. (A.L.B.)

  2. Wind Energy

    Energy Technology Data Exchange (ETDEWEB)

    Ganley, Jason; Zhang, Jie; Hodge, Bri-Mathias

    2016-03-15

    Wind energy is a variable and uncertain renewable resource that has long been used to produce mechanical work, and has developed into a large producer of global electricity needs. As renewable sources of energy and feedstocks become more important globally to produce sustainable products, many different processes have started adopting wind power as an energy source. Many times this is through a conversion to hydrogen through electrolysis that allows for a more continuous process input. Other important pathways include methanol and ammonia. As the demand for sustainable products and production pathways increases, and wind power capital costs decrease, the role of wind power in chemical and energy production seems poised to increase significantly.

  3. Basic research for nuclear energy. y Study on the nuclear materials technology

    Energy Technology Data Exchange (ETDEWEB)

    Kuk, I. H.; Lee, H. S.; Jeong, Y. H.; Sung, K. W.; Han, J. H.; Lee, J. T.; Lee, H. K.; Kim, S. J.; Kang, H. S.; An, D. H.; Kim, K. R.; Park, S. D.; Han, C. H.; Jung, M. K.; Oh, Y. J.; Kim, K. H.; Kim, S. H.; Back, J. H.; Kim, C. H.; Lim, K. S.; Kim, Y. Y.; Na, J. W.; Ku, J. H.; Lee, D. H.

    1996-12-01

    A study on the nuclear materials technologies which are necessary to establish the base for alloy development was performed. - The feasibility study on the application of Zircaloy scrap waste for hydrogen storage - The development of metal hydride battery for energy storage system - The establishment of transmission electron microscopy database for nuclear materials - The basic technology for the development of cladding materials for high burnup - The water chemistry technology for secondary system pH control and the photocatalysis technology for decomposition and removal of organics. - Improvement of primary component integrity of PWR by Zinc injection. (author). 175 refs., 58 tabs., 262 figs.

  4. Discretization of Continuous Frame

    Indian Academy of Sciences (India)

    A Fattahi; H Javanshiri

    2012-05-01

    In this paper we consider the notion of continuous frame of subspaces and define a new concept of continuous frame, entitled continuous atomic resolution of identity, for arbitrary Hilbert space $\\mathcal{H}$ which has a countable reconstruction formula. Among the other results, we characterize the relationship between this new concept and other known continuous frames. Finally, we state and prove the assertions of the stability of perturbation in this concept.

  5. 基于连续能量蒙特卡罗方法的均匀化群常数计算%Continuous energy Monte Carlo method based homogenization multi-group constants calculation

    Institute of Scientific and Technical Information of China (English)

    李满仓; 王侃; 姚栋

    2012-01-01

    两步法反应堆物理计算流程中,组件均匀化群常数显著影响堆芯计算精度.相比确定论方法,连续能量蒙特卡罗方法均匀化精确描述各种几何构型栅格,避免繁琐共振自屏计算,保留更多连续能量信息,不仅产生的群常数更精确,而且普适性也更强.作为实现连续能量蒙特卡罗组件均匀化的第一步,本文应用径迹长度方法统计计算一般群截面和群常数,提出并使用散射事件方法获得不能直接应用确定论方法计算群间散射截面和高阶勒让德系数,应用P1截面计算扩散系数.为还原两步法计算流程中组件在堆芯的临界状态,本文应用BN理论对均匀化群常数进行泄漏修正.在4种类型组件和简化压水堆堆芯上数值验证蒙特卡罗均匀化群常数.验证结果表明:连续能量蒙特卡罗方法组件均匀化群常数具有良好几何适应性,显著提高堆芯计算精度.%The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P1 cross sections were used to calculate the diffusion

  6. Continuous Markovian Logics

    DEFF Research Database (Denmark)

    Mardare, Radu Iulian; Cardelli, Luca; Larsen, Kim Guldstrand

    2012-01-01

    Continuous Markovian Logic (CML) is a multimodal logic that expresses quantitative and qualitative properties of continuous-time labelled Markov processes with arbitrary (analytic) state-spaces, henceforth called continuous Markov processes (CMPs). The modalities of CML evaluate the rates of the ...

  7. Plants under continuous light

    NARCIS (Netherlands)

    Velez Ramirez, A.I.; Ieperen, van W.; Vreugdenhill, D.; Millenaar, F.F.

    2011-01-01

    Continuous light is an essential tool for understanding the plant circadian clock. Additionally, continuous light might increase greenhouse food production. However, using continuous light in research and practice has its challenges. For instance, most of the circadian clock-oriented experiments wer

  8. An investigation into the origen of the interference generated during the measurement of the reactivity in a high burn-up reactor core%高燃耗堆芯反应性测量的干扰源研究

    Institute of Scientific and Technical Information of China (English)

    陈雄月; 吕大军; 裘希春; 韩承慈; 夏应军; 邓朝平; 张仲元

    2012-01-01

    回顾了1980年9月实验前,在反应堆噪声分析领域的技术发展概况.展示了在实验动力堆燃耗末、卸料前,用双探测器互相关频谱分析法(CCFS)测得的一组数据;和经过离线去本底拟合计算后,获得的动力学参数测量结果:αc=(144.57±2.09)s-1.介绍了数据获取过程中出现的异常情况;离线处理的方法;本底谱选定;拟合计算程序;计算结果和结论.还简要介绍了干扰源的来源及其强度计算概况.数据处理结果证明:在长期燃耗后的堆芯上应用噪声分析法,除了要克服大γ场的干扰外,还要严格消除本底中子场产生的不相关噪声干扰.%After a general review for the technical development before 1980's in the area of nuclear reactor noise analysis,a reactor dynamic parameter,ac = (144. 57 + 2. 09)s~1 , obtained through off-line background processing, is shown. The processed data is measured through double- detector cross correlation frequency spectral analysis (CCFS) for the experimental nuclear power reactor at the burn-up end in sept. 1980. This paper also presents the abnormal situations for data acquisition, the off-line data processing method,the background spectra selection for data processing and the program for the least-squares fit calculation. Here also explains how neutron background is generated and how its strength is calculated. This verifies the fact that after a long-term burn-up run, large y field must be suppressed and also more attention must be paid to the uncorrelated neutron noise from the fuel burn-up.

  9. A Phased Development of Breed-and-Burn Reactors for Enhanced Nuclear Energy Sustainability

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available Several options for designing fast reactors to operate in the Breed-and-Burn (B&B mode are compared and a strategy is outlined for early introduction of B&B reactors followed by a gradual increase in the fuel utilization of such reactors. In the first phase the fast reactor core will consist of a subcritical B&B blanket driven by a relatively small critical seed. As the required discharge burnup/radiation-damage to both driver and blanket fuel had already been proven, and as the depleted uranium fueled B&B blanket could generate close to 2/3 of the core power and will have very low fuel cycle cost, the deployment of such fast reactors could start in the near future. The second phase consists of deploying self-sustaining stationary wave B&B reactors. It will require development of fuel technology that could withstand peak burnups of ~30% and peak radiation damage to the cladding of ~550 dpa. The third phase requires development of a fuel reconditioning technology that will enable using the fuel up to an average burnup of ~50%—the upper bound permitted by neutron balance considerations when most of the fission products are not separated from the fuel. The increase in the uranium ore utilization relative to that provided by contemporary power reactors is estimated to be 20, 40 and 100 folds for, respectively, phase 1, 2 and 3. The energy value of the depleted uranium stockpiles (“waste” accumulated in the US is equivalent to, when used in the B&B reactors, up to 20 centuries of the total 2010 USA supply of electricity. Therefore, a successful development of B&B reactors could provide a great measure of energy sustainability and cost stability.

  10. Energy and technology review

    Energy Technology Data Exchange (ETDEWEB)

    Selden, R.W.

    1977-05-01

    Topics covered include: geothermal energy development at LLL, energy conversion engineering, continuing education at LLL, and the Western states uranium resource survey. Separate abstracts were prepared for 3 sections. (MCG)

  11. Continuity in Discrete Sets

    CERN Document Server

    Burgin, Mark

    2010-01-01

    Continuous models used in physics and other areas of mathematics applications become discrete when they are computerized, e.g., utilized for computations. Besides, computers are controlling processes in discrete spaces, such as films and television programs. At the same time, continuous models that are in the background of discrete representations use mathematical technology developed for continuous media. The most important example of such a technology is calculus, which is so useful in physics and other sciences. The main goal of this paper is to synthesize continuous features and powerful technology of the classical calculus with the discrete approach of numerical mathematics and computational physics. To do this, we further develop the theory of fuzzy continuous functions and apply this theory to functions defined on discrete sets. The main interest is the classical Intermediate Value theorem. Although the result of this theorem is completely based on continuity, utilization of a relaxed version of contin...

  12. On barely continuous functions

    Directory of Open Access Journals (Sweden)

    Richard Stephens

    1988-01-01

    Full Text Available The term barely continuous is a topological generalization of Baire-1 according to F. Gerlits of the Mathematical Institute of the Hungarian Academy of Sciences, and thus worthy of further study. This paper compares barely continuous functions and continuous functions on an elementary level. Knowing how the continuity of the identity function between topologies on a given set yields the lattice structure for those topologies, the barely continuity of the identity function between topologies on a given set is investigated and used to add to the structure of that lattice. Included are certain sublattices generated by the barely continuity of the identity function between those topologies. Much attention is given to topologies on finite sets.

  13. Cutting Out Continuations

    DEFF Research Database (Denmark)

    Bahr, Patrick; Hutton, Graham

    2016-01-01

    In the field of program transformation, one often transforms programs into continuation-passing style to make their flow of control explicit, and then immediately removes the resulting continuations using defunctionalisation to make the programs first-order. In this article, we show how these two...... transformations can be fused together into a single transformation step that cuts out the need to first introduce and then eliminate continuations. Our approach is calculational, uses standard equational reasoning techniques, and is widely applicable....

  14. Transitions in Energy Use

    OpenAIRE

    Grubler, A.

    2004-01-01

    Patterns of energy use have changed dramatically since the onset of the industrial revolution in terms of both energy quantities and energy quality. These changing patterns of energy use, where energy quantities and quality interact in numerous important ways, are referred to in this article as energy transitions and are described from a historical perspective as well as through future scenarios. Far from being completed, many of these transitions are continuing to unfold in industrial and de...

  15. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  16. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

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

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  17. Hanford year 2000 Business Continuity Plan

    Energy Technology Data Exchange (ETDEWEB)

    ROGGENKAMP, S.L.

    1999-11-01

    The goal of Department of Energy Richland Operations (DOE-RL) Year 2000 (Y2K) effort is to ensure that the Hanford site successfully continues its mission as we approach and enter the 21th century. The Y2K Business Continuity Planning process provides a structured approach to identify Y2K risks to the site and to mitigate these risks through Y2K Contingency Planning, ''Zero-Day'' Transition Planning and Emergency Preparedness. This document defines the responsibilities, processes and plans for Hanford's Y2K Business Continuity. It identifies proposed business continuity drills, tentative schedule and milestones.

  18. Let Continuous Outcome Variables Remain Continuous

    Directory of Open Access Journals (Sweden)

    Enayatollah Bakhshi

    2012-01-01

    Full Text Available The complementary log-log is an alternative to logistic model. In many areas of research, the outcome data are continuous. We aim to provide a procedure that allows the researcher to estimate the coefficients of the complementary log-log model without dichotomizing and without loss of information. We show that the sample size required for a specific power of the proposed approach is substantially smaller than the dichotomizing method. We find that estimators derived from proposed method are consistently more efficient than dichotomizing method. To illustrate the use of proposed method, we employ the data arising from the NHSI.

  19. Black holes by analytic continuation

    CERN Document Server

    Amati, Daniele

    1997-01-01

    In the context of a two-dimensional exactly solvable model, the dynamics of quantum black holes is obtained by analytically continuing the description of the regime where no black hole is formed. The resulting spectrum of outgoing radiation departs from the one predicted by the Hawking model in the region where the outgoing modes arise from the horizon with Planck-order frequencies. This occurs early in the evaporation process, and the resulting physical picture is unconventional. The theory predicts that black holes will only radiate out an energy of Planck mass order, stabilizing after a transitory period. The continuation from a regime without black hole formation --accessible in the 1+1 gravity theory considered-- is implicit in an S matrix approach and provides in this way a possible solution to the problem of information loss.

  20. Monolithic Continuous-Flow Bioreactors

    Science.gov (United States)

    Stephanopoulos, Gregory; Kornfield, Julia A.; Voecks, Gerald A.

    1993-01-01

    Monolithic ceramic matrices containing many small flow passages useful as continuous-flow bioreactors. Ceramic matrix containing passages made by extruding and firing suitable ceramic. Pores in matrix provide attachment medium for film of cells and allow free movement of solution. Material one not toxic to micro-organisms grown in reactor. In reactor, liquid nutrients flow over, and liquid reaction products flow from, cell culture immobilized in one set of channels while oxygen flows to, and gaseous reaction products flow from, culture in adjacent set of passages. Cells live on inner surfaces containing flowing nutrient and in pores of walls of passages. Ready access to nutrients and oxygen in channels. They generate continuous high yield characteristic of immobilized cells, without large expenditure of energy otherwise incurred if necessary to pump nutrient solution through dense biomass as in bioreactors of other types.

  1. Residential Continuing Education.

    Science.gov (United States)

    Houle, Cyril O.

    The theme of this discursive essay is residential continuing education: its definition, its development along somewhat different lines in Europe and in America, and its practice in university centers in the United States. Continuing education includes any learning or teaching program that is based on the assumptions that the learners have studied…

  2. On continued fraction algorithms

    NARCIS (Netherlands)

    Smeets, Ionica

    2010-01-01

    Is there a good continued fraction approximation between every two bad ones? What is the entropy of the natural extension for alpha-Rosen fractions? How do you find multi-dimensional continued fractions with a guaranteed quality in polynomial time? These, and many more, questions are answered in thi

  3. On continued fraction algorithms

    NARCIS (Netherlands)

    Smeets, Ionica

    2010-01-01

    Is there a good continued fraction approximation between every two bad ones? What is the entropy of the natural extension for alpha-Rosen fractions? How do you find multi-dimensional continued fractions with a guaranteed quality in polynomial time? These, and many more, questions are answered in thi

  4. Development of the MCNPX depletion capability: A Monte Carlo linked depletion method that automates the coupling between MCNPX and CINDER90 for high fidelity burnup calculations

    Science.gov (United States)

    Fensin, Michael Lorne

    Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established

  5. LANL continuity of operations plan

    Energy Technology Data Exchange (ETDEWEB)

    Senutovitch, Diane M [Los Alamos National Laboratory

    2010-12-22

    The Los Alamos National Laboratory (LANL) is a premier national security research institution, delivering scientific and engineering solutions for the nation's most crucial and complex problems. Our primary responsibility is to ensure the safety, security, and reliability of the nation's nuclear stockpile. LANL emphasizes worker safety, effective operational safeguards and security, and environmental stewardship, outstanding science remains the foundation of work at the Laboratory. In addition to supporting the Laboratory's core national security mission, our work advances bioscience, chemistry, computer science, earth and environmental sciences, materials science, and physics disciplines. To accomplish LANL's mission, we must ensure that the Laboratory EFs continue to be performed during a continuity event, including localized acts of nature, accidents, technological or attack-related emergencies, and pandemic or epidemic events. The LANL Continuity of Operations (COOP) Plan documents the overall LANL COOP Program and provides the operational framework to implement continuity policies, requirements, and responsibilities at LANL, as required by DOE 0 150.1, Continuity Programs, May 2008. LANL must maintain its ability to perform the nation's PMEFs, which are: (1) maintain the safety and security of nuclear materials in the DOE Complex at fixed sites and in transit; (2) respond to a nuclear incident, both domestically and internationally, caused by terrorist activity, natural disaster, or accident, including mobilizing the resources to support these efforts; and (3) support the nation's energy infrastructure. This plan supports Continuity of Operations for Los Alamos National Laboratory (LANL). This plan issues LANL policy as directed by the DOE 0 150.1, Continuity Programs, and provides direction for the orderly continuation of LANL EFs for 30 days of closure or 60 days for a pandemic/epidemic event. Initiation of COOP operations may

  6. Microeconomic and macroeconomic analysis of the cost and benefit effects of the continued development of renewable energies in the German electricity and heat market. Stock-taking and assessment of current methods of quantifying the cost and benefit effects of the continued development of renewable energies in the electricity and heat sector. Work package 1; Einzel- und gesamtwirtschaftliche Analyse von Kosten- und Nutzenwirkungen des Ausbaus Erneuerbarer Energien im deutschen Strom- und Waermemarkt. Bestandsaufnahme und Bewertung vorliegender Ansaetze zur Quantifizierung der Kosten-Nutzen-Wirkungen des Ausbaus Erneuerbarer Energien im Strom- und Waermebereich. Arbeitspaket 1

    Energy Technology Data Exchange (ETDEWEB)

    Breitschopf, Barbara; Klobasa, Marian; Sensfuss, Frank [Fraunhofer-Institut fuer System- und Innovationsforschung (ISI), Karlsruhe (DE)] (and others)

    2010-03-15

    The present study forms the first part of an extensive project dedicated to an analysis of the cost and benefit effects of the continued development of renewable energies in the electricity and heat sectors. The project's objective is to establish a scientifically substantiated, generally accepted basis for a comprehensive assessment of the use and continued development of renewable energies up to the present. For this purpose the following report presents an overview and discussion of the approaches and methods used in the past for determining such effects. While some effects have already been quantified and discussed at length, others have largely been ignored or only considered in qualitative terms. The study particularly gives impetus to the analysis of cost and benefit effects in the heat sector, whereas in regard to the electricity sector it largely focuses on a thorough evaluation of studies already performed. All studies under review are categorised within a conceptual framework for the purpose of arriving at an overall assessment of effects without double counting or gaps. Effects are classified as follows, depending on the category in question: costs and benefits of renewable energy technologies; burdening of and relief to market participants; and effects on value creation and employment. Wherever possible, effects are quantified in monetary terms stating the object (e.g. electricity from renewable energy) and range of analysis in each case. The analysis horizon has initially been defined such that it captures the present situation (2007 and possibly 2008). The second part of the project covers follow-on or in-depth studies aimed at improving estimates and representations of current and future effects (in the years from 2020 to 2030).

  7. Continuous Markovian Logics

    DEFF Research Database (Denmark)

    Mardare, Radu Iulian; Cardelli, Luca; Larsen, Kim Guldstrand

    2012-01-01

    Continuous Markovian Logic (CML) is a multimodal logic that expresses quantitative and qualitative properties of continuous-time labelled Markov processes with arbitrary (analytic) state-spaces, henceforth called continuous Markov processes (CMPs). The modalities of CML evaluate the rates...... characterizes stochastic bisimilarity and it supports the definition of a quantified extension of the satisfiability relation that measures the "compatibility" between a model and a property. In this context, the metaproperties allows us to prove two robustness theorems for the logic stating that one can...

  8. The Q-factor of a continuous-wave laser

    NARCIS (Netherlands)

    Eichhorn, M.; Pollnau, Markus

    We define the finite Q-factor of a continuous-wave lasing resonator as the energy of coherent photons stored in the resonator at a given time over the energy of these coherent photons lost per oscillation cycle.

  9. Nocturnal continuous glucose monitoring

    DEFF Research Database (Denmark)

    Bay, Christiane; Kristensen, Peter Lommer; Pedersen-Bjergaard, Ulrik;

    2013-01-01

    Abstract Background: A reliable method to detect biochemical nocturnal hypoglycemia is highly needed, especially in patients with recurrent severe hypoglycemia. We evaluated reliability of nocturnal continuous glucose monitoring (CGM) in patients with type 1 diabetes at high risk of severe...

  10. Continuous ethanol fermentors

    Energy Technology Data Exchange (ETDEWEB)

    1983-08-03

    A continuous EtOH fermentor was developed. In the 1st stage of the fermentor, EtOH fermentation medium is contacted with an EtOH-producing bacterium (e.g. Zymomonas mobilis) attached to a carrier material (e.g., vermiculite powder) and with brewers' bottom yeast in the 2nd stage. This system does not require any special cell separator for continuous operation.

  11. Continuous Time Model Estimation

    OpenAIRE

    Carl Chiarella; Shenhuai Gao

    2004-01-01

    This paper introduces an easy to follow method for continuous time model estimation. It serves as an introduction on how to convert a state space model from continuous time to discrete time, how to decompose a hybrid stochastic model into a trend model plus a noise model, how to estimate the trend model by simulation, and how to calculate standard errors from estimation of the noise model. It also discusses the numerical difficulties involved in discrete time models that bring about the unit ...

  12. Continuous parallel coordinates.

    Science.gov (United States)

    Heinrich, Julian; Weiskopf, Daniel

    2009-01-01

    Typical scientific data is represented on a grid with appropriate interpolation or approximation schemes,defined on a continuous domain. The visualization of such data in parallel coordinates may reveal patterns latently contained in the data and thus can improve the understanding of multidimensional relations. In this paper, we adopt the concept of continuous scatterplots for the visualization of spatially continuous input data to derive a density model for parallel coordinates. Based on the point-line duality between scatterplots and parallel coordinates, we propose a mathematical model that maps density from a continuous scatterplot to parallel coordinates and present different algorithms for both numerical and analytical computation of the resulting density field. In addition, we show how the 2-D model can be used to successively construct continuous parallel coordinates with an arbitrary number of dimensions. Since continuous parallel coordinates interpolate data values within grid cells, a scalable and dense visualization is achieved, which will be demonstrated for typical multi-variate scientific data.

  13. On a connection between the limit set of the Moebius-Klein transformation, periodic continued fractions, El Naschie's topological theory of high energy par